ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 102

DOE Review Release 12/04/2009

Document Title:

An Exposure Matrix for Linde Ceramics Plant 

(Including Tonawanda Laboratory)

Document Number: ORAUT-TKBS-0025

Revision: 01 PC-1

Effective Date: 11/10/2009

Type of Document: TBD

Supersedes: Revision 01

Subject Expert(s): Gerald R. Davidson

Site Expert(s): N/A

Approval: Signature on File Approval Date: 11/05/2009

Mutty Sharfi, Document Owner

Concurrence: Signature on File Concurrence Date: 11/04/2009

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 11/04/2009

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 11/04/2009

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 11/10/2009

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 2 of 102

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

05/31/2005 00 New Technical Basis Document: Basis for the Development of an 

Exposure Matrix for Linde Air Products, Tonawanda, New York and 

Buffalo, New York. First approved issue. Initiated by Cindy Bloom.

01/19/2006 00 PC-1 Modified to incorporate comments from OCAS, a Worker Outreach 

meeting comment and minor editing. The Introduction on page 10 

was updated. Dates were changed in Section 3.0 to incorporate the 

decision to designate Linde Ceramics workers, who received internal 

exposures prior to 11/01/1947, as a Special Exposure Cohort. 

Moved Table 12, an internal exposure table out of Section 4.0. 

Changes were made to page 23 in Section 3, page 34 in Table 7, 

page 35 in Table 8, page 38 in Table 11, and Table 12 on page 39. 

Approved page change revision. No sections were deleted. Training 

required: As determined by the Task Manager. Initiated by Cindy W. 

Bloom. 

Signature on File 01/18/2006

Cindy W. Bloom, TBD Team Leader

Signature on File 01/18/2006

Judson L. Kenoyer, Task 3 Manager

Signature on File 01/18/2006

Kate Kimpan, Project Director

James W. Neton, Associate Director for Science

Signature on File 01/19/2006

11/04/2008 01 Approved revision to change from a page change revision (Rev 00 

PC-2-B) to a total rewrite (Rev 01-A) as a result of formal NIOSH 

review. Revised to incorporate: (1) change in facility designation, (2) 

DOL interpretation of applicability of residual period to Ceramics 

Plant, (3) and resolution of Advisory Board Working Group 

comments, (4) clarified the implementation instructions for 

SEC00044 for the period October 1, 1942 through October 31, 1947. 

Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Task Manager. Initiated by Joseph S. Guido.

11/10/2009 01 PC-1 Approved page change revision to incorporate additional guidance 

on the use of Attachment E of the document. This change 

constitutes the resolution of Advisory Board Working Group concern 

and clarifies how to proceed in the event of credible indication of 

potential exposure to contaminated burlap bags that were used for 

transport of ore at Linde Ceramics. Changes were made to page 37 

in Section 4.0. No sections were deleted. No changes occurred as a 

result of formal internal review. Incorporates formal NIOSH review 

comments. Training required: As determined by the Objective 

Manager. Initiated by Mutty Sharfi. 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 3 of 102

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

7

1.0 Introduction .............................................................................................................................. 9

1.1 Purpose......................................................................................................................10

1.2 Scope.........................................................................................................................11

2.0 Site Description and Operational History................................................................................ 12

2.1 Facilities .....................................................................................................................12

2.1.1 Buildings A and B............................................................................................ 14

2.1.2 Building 14 (Tonawanda Laboratory) .............................................................. 15

2.1.3 Building 30 (Ceramics Plant)........................................................................... 15

2.1.4 Building 31 (Ceramics Plant)........................................................................... 15

2.1.5 Building 37 (Ceramics Plant)........................................................................... 16

2.1.6 Building 38 (Ceramics Plant)........................................................................... 16

2.2 Source Terms for African and Domestic Uranium Ores ..............................................16

2.3 Activities .....................................................................................................................18

2.3.1 Uranium Research and Development ............................................................. 19

2.3.2 Uranium Production ........................................................................................ 19

2.3.2.1 Step I (Uranium Ore to U3O8)............................................................ 20

2.3.2.2 Step II (U3O8 to UO2) ........................................................................ 21

2.3.2.3 Step III (UO2 to UF4) ......................................................................... 22

2.3.3 Nickel Production............................................................................................ 22

2.3.4 Other Radiological Activities............................................................................ 23

2.4 Protective Measures...................................................................................................23

2.5 Personnel, Job Categories, And Work hours ..............................................................24

2.6 Decontamination During the MED/AEC Contract Period.............................................26

2.7 Post-MED/AEC Operations ........................................................................................27

3.0 Estimation of Internal Exposure, 1947 to July 7, 1954............................................................ 28

3.1 Estimation of Particulate Intakes.................................................................................28

3.2 Uranium......................................................................................................................29

3.2.1 Uranium Urinalysis Data ................................................................................. 29

3.3 Radium.......................................................................................................................31

3.4 Uranium Progeny........................................................................................................31

3.5 Radon.........................................................................................................................32

3.5.1 Ceramics Plant Radon Exposures................................................................... 32

3.5.2 Tonawanda Laboratory Radon Exposures ...................................................... 33

3.6 Inhalation Intake Estimates of Radioactive Particulates at Tonawanda Laboratory.....34

3.7 Ingestion Intake Estimates at Tonawanda Laboratory ................................................35

3.8 Occupational Internal Dose Reconstruction Assumptions and Summary....................35

4.0 Estimation of External Exposure, 1942 to July 7, 1954........................................................... 37

4.1 Ceramics Plant Beta and Gamma Exposure...............................................................37

4.1.1 Preproduction, 1942 to 1943........................................................................... 37

4.1.2 Production, 1943 to 1946 ................................................................................ 39

4.1.2.1 Beta.................................................................................................. 39

4.1.2.1.1 Estimation of Worker Doses ............................................ 41

4.1.2.1.2 Outdoor Dose Rate ......................................................... 43

4.1.2.1.3 Categories....................................................................... 43
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4.1.2.2 Gamma ............................................................................................ 44

4.1.2.2.1 Estimation of Worker Doses ............................................ 48

4.1.2.2.2 Outdoor Exposure Rate................................................... 51

4.1.3 Standby, 1946 to 1947.................................................................................... 52

4.1.4 Rehabilitation and Production, 1947 to 1949................................................... 52

4.1.4.1 Beta.................................................................................................. 52

4.1.4.2 Gamma ............................................................................................ 56

4.1.5 Cleanup, 1949 to July 7, 1954......................................................................... 57

4.2 Tonawanda Laboratory Beta and Gamma Exposure ..................................................59

4.3 Neutron Exposure.......................................................................................................59

4.4 External Dose Reconstruction Summary, October 1, 1942, to July 7, 1954 ................62

5.0 Occupational Medical Exposure............................................................................................. 66

5.1 Ceramics Plant Employees.........................................................................................66

5.1.1 Chest X-rays ................................................................................................... 66

5.1.2 Pelvis/Lumbar Spine X-rays ............................................................................ 68

5.1.3 X-Ray Dose Reconstruction Guidelines .......................................................... 70

5.2 Tonawanda Laboratory Employees ............................................................................70

6.0 Estimation of Exposures from Residual Contamination after 1953 ......................................... 72

6.1 Internal Exposure .......................................................................................................72

6.2 External Beta and Gamma Exposure..........................................................................73

6.3 Neutron Exposure.......................................................................................................74

7.0 Attributions and Annotations .................................................................................................. 74

References ........................................................................................................................................ 75

Glossary ............................................................................................................................................ 86

ATTACHMENT A, BETA RADIATION .............................................................................................. 87

ATTACHMENT B, CODES AND SPECIAL TERMINOLOGY............................................................91

ATTACHMENT C, DATA SOURCES ON URANIUM PROGENY CONCENTRATIONS IN 

LINDE MATERIALS ............................................................................................92

ATTACHMENT D, LINDE URANIUM COWORKER ASSESSMENT FOR NOVEMBER 1947

TO JANUARY 1950.............................................................................................93

ATTACHMENT E, FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM 

URANIUM ORE BAGS ON THE SITE DURING THE POSTOPERATIONS 

PERIOD ..............................................................................................................98
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 5 of 102

LIST OF TABLES

TABLE TITLE

2-1 Key dates...............................................................................................................................

PAGE

13

2-2 Department codes for various worker activities ...................................................................... 24

2-3 Worker information................................................................................................................. 25

2-4 Ceramics Plant assumed work schedule (including lunch and breaks), 1942 to 

July 7, 1954............................................................................................................................26

3-1 Chronic intake rate for Type M uranium ................................................................................. 31

3-2 Chronic intake rate for Type S uranium.................................................................................. 31

3-3 Progeny to uranium ratios ...................................................................................................... 32

3-4 Measured radon concentrations during Ceramics Plant African ore processing ..................... 33

3-5 Ceramics Plant worker radon exposure rates, 1947 to 1954 .................................................. 33

3-6 Tonawanda Laboratory radon exposure rates, 1947 to 1953 ................................................. 34

3-7 Assumed airborne concentrations used to estimate intakes at the Tonawanda 

Laboratory..............................................................................................................................34

3-8 Internal exposure summary for the Ceramics Plant, November 1, 1947, to July 7, 1954 

(based on Type M uranium) ................................................................................................... 35

3-9 Internal exposure summary for the Ceramics Plant, November 1, 1947, to July 7, 1954 

(based on Type S uranium).................................................................................................... 36

3-10 Internal exposure summary for the Ceramics Plant, November 1, 1947, to July 7, 1954 ........ 36

3-11 Internal exposure summary for the Tonawanda Laboratory, November 1, 1947, to 

December 31, 1953................................................................................................................36

4-1 Floor and wall radiation in Building 30 before and after the 1949 to 1950 

decontamination.....................................................................................................................38

4-2 Building 30 floor radiation levels in 1976 ................................................................................ 38

4-3 Preproduction beta and gamma radiation levels..................................................................... 39

4-4 Beta radiation exposure rates measured at Linde Ceramics, January 23, 1944 ..................... 40

4-5 Dependence of beta dose rate on ore type ............................................................................ 42

4-6 Assigned beta dose rates, 1943 to 1946 ................................................................................ 43

4-7 Job categories for Step I gamma dose analysis, 1943 to 1946............................................... 45

4-8 Step I gamma exposure data for African ore processing, January 31 to November 12, 

1944.......................................................................................................................................46

4-9 Step I gamma exposure rate parameters for African ore processing, January 31 to 

November 12, 1944................................................................................................................46

4-10 Step I gamma exposure data for cleanup and domestic ore processing, November 13, 

1944, to February 26, 1945.................................................................................................... 47

4-11 Step I gamma exposure rate parameters for cleanup and domestic ore processing, 

November 13, 1944, to February 26, 1945............................................................................. 47

4-12 Estimated maximum exposure rates in selected Building 30 facilities due to 30-by 30-

by 12-ft pile of 8% ore in southeast corner ............................................................................. 51

4-13 Ceramics Plant beta and gamma radiation rates during standby............................................ 52

4-14 Job categories for 1947 to 1949 Step III film badge analysis.................................................. 53

4-15 Step III beta dose data, 1947 to 1949..................................................................................... 54

4-16 Beta dose rates for Step III, 1947 to 1949.............................................................................. 55

4-17 Step III gamma exposure data, 1947 to 1949......................................................................... 56

4-18 Gamma exposure rates for Step III, 1947 to 1949.................................................................. 57

4-19 Beta and gamma radiation measured during equipment removal, 1948................................. 58

4-20 External dose reconstruction parameters for cleanup............................................................ 59

4-21 Tonawanda Laboratory beta and gamma radiation, 1942 to 1953.......................................... 60

4-22 Natural uranium per gram dose rates at 1 ft ........................................................................... 60

4-23 Neutron dose rates and annual doses from the alpha-neutron sources.................................. 62
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 6 of 102

4-24 Summary of annual external exposure from AWE operations, 1942 to 1953.......................... 63

4-25 Ceramics Plant worker beta and gamma external exposure categories for production 

periods...................................................................................................................................65

6-1 Highest monthly outdoor airborne radionuclide concentrations at Linde Tonawanda 

site, 2000 to 2004 ..................................................................................................................72

6-2 Annual internal and external exposure to residual radioactivity .............................................. 74

A-1 Maximum beta surface dose rates from various uranium-containing materials....................... 87

A-2 Electron energy released by refined uranium ......................................................................... 87

A-3 Electron energy released by unrefined natural uranium ore ................................................... 88

A-4 Estimated attenuation of beta radiation with distance............................................................. 89

A-5 Uranium beta dose reduction factors for apparel.................................................................... 89

A-6 Variation of dose rate with body location for Step III process operations ................................ 90

B-1 Codes and special terminology .............................................................................................. 91

D-1 Linde data sources in the Site Research Database................................................................ 94

D-2 Uranium urinary excretion data .............................................................................................. 95

D-3 Chronic intake rates for Type M 234U ...................................................................................... 96

D-4 Chronic intake rates for Type S 234U....................................................................................... 96

LIST OF FIGURES

FIGURE TITLE

2-1 Linde Ceramics Plant and Tonawanda Laboratory buildings on LAPC Tonawanda site ........

PAGE

14

2-2 Building 30 layout...................................................................................................................16

2-3 Building 30 layout...................................................................................................................17

2-4 Building 38 layout...................................................................................................................18

4-1 Fit to gamma exposure data for a 4- by 5- by 20-ft pile of 8% African pitchblende ore ........... 49

4-2 Calculated exposure rates and analytic fit for a 30- by 30- by 12-ft pile of 8% African 

pitchblende ore ......................................................................................................................50

D-1 Type M 50th-percentile value for November 1, 1947, to January 27, 1950............................. 96

D-2 Type S 50th-percentile value for November 1, 1947, to January 27, 1950 ............................. 96

D-3 Type M 84th-percentile value for November 1, 1947, to January 27, 1950............................. 97

D-4 Type S 84th-percentile value for November 1, 1947, to January 27, 1950 ............................. 97
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 7 of 102

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

CFR Code of Federal Regulations

cm centimeter

d day

D&D decontamination and decommissioning

DOE U.S. Department of Energy

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

FUSRAP Formerly Utilized Sites Remedial Action Program

GM geometric mean

GSD geometric standard deviation

hr hour

ICRP International Commission on Radiological Protection

keV kiloelectron-volt, 1,000 electron-volts

L liter

lb pound

LAPC Linde Air Products Company

LOD limit of detection

LOOW Lake Ontario Ordnance Works

MAC maximum allowable concentration

m meter

MED Manhattan Engineer District

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mm millimeter

mr radiation exposure unit used by MED in the 1940s; interpreted here as millirad for 

photons and millirem for beta radiation

mR milliroentgen

mrad millirad

mrem millirem

mrep millirep

MT metric ton (1,000 kg)

NIOSH National Institute for Occupational Safety and Health

NYOO New York Operations Office 

OCAS Office of Compensation Analysis and Support

ORNL Oak Ridge National Laboratory

ORAU Oak Ridge Associated Universities
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 8 of 102

PAEC potential alpha energy concentration (a measure of concentration of radon progeny)

PAEE potential alpha energy exposure (a measure of exposure to radon progeny)

pCi picocuries

PL preferred level

POC probability of causation

R roentgen

R&D research and development

s second

SC&A S. Cohen & Associates

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

t ton (2,000 lb)

U.S.C. United States Code

USACE U.S. Army Corps of Engineers

USBM U.S. Bureau of Mines

wd workday

wk week

WL working level

WLM working level-month

yr year

α alpha particle

μg microgram

μm micrometer

μR microroentgen

°C degrees Celsius

°F degrees Fahrenheit

§ section or sections
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 9 of 102

1.0

Technical basis documents (TBDs) and site profile documents are not official determinations made by 

the National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

INTRODUCTION

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds on 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2007a):

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007a).

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 10 of 102

Under EEOICPA, employment at an AWE facility is categorized as either (1) during the DOE contract 

period (i.e., when the AWE was processing or producing material that emitted radiation and was used 

in the production of an atomic weapon), or (2) during the residual contamination period (i.e., periods 

that NIOSH has determined there is the potential for significant residual contamination after the period 

in which weapons-related production occurred). For contract period employment, all occupationally 

derived radiation exposures at covered facilities must be included in dose reconstructions. This 

includes radiation exposure related to the Naval Nuclear Propulsion Program and any radiation 

exposure received from the production of commercial radioactive products that were concurrently 

manufactured by the AWE facility during the covered period. NIOSH does not consider the following 

exposures to be occupationally derived (NIOSH 2007a):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

For employment during the residual contamination period, only the radiation exposures defined in 42 

U.S.C. § 7384n(c)(4) [i.e., radiation doses received from DOE-related work] must be included in dose 

reconstructions. Doses from medical X-rays are not reconstructed during the residual contamination 

period (NIOSH 2007a). It should be noted that under subparagraph A of 42 U.S.C. § 7384n(c)(4), 

radiation associated with the Naval Nuclear Propulsion Program is specifically excluded from the 

employee’s radiation dose. This exclusion only applies to those AWE employees who worked during 

the residual contamination period. Also, under subparagraph B of 42 U.S.C. § 7384n(c)(4), radiation 

from a source not covered by subparagraph A that is not distinguishable through reliable 

documentation from radiation that is covered by subparagraph A is considered part of the employee’s 

radiation dose. This site profile covers only exposures resulting from nuclear weapons-related work.

Exposures resulting from non-weapons-related work, if applicable, will be covered elsewhere.

1.1 PURPOSE

This site profile document provides an exposure matrix for workers at the Tonawanda Laboratory and 

Linde Ceramics Plant facilities of the Linde Air Products Company (LAPC) in Tonawanda, New York.

NIOSH has determined, and the Secretary, Health and Human Services has concurred, that it is not 

feasible to reconstruct internal radiation dose for: “Atomic weapons employees who worked at the 

Linde Ceramics Plant from October 1, 1942, through October 31, 1947, and who were employed for a 

number of work days aggregating at least 250 work days either solely under this employment or in 

combination with work days occurring within the parameters (excluding aggregate work day 

requirements) established for other classes of employees included in the SEC” (HHS 2005). 

Subsequent correspondence (Elliott 2006) confirms that the Tonawanda Laboratory (as well as all 

other buildings on the Linde Site) are included in this class designation (cohort). Reconstruction of 

external exposure (including medical X-ray examinations) has been determined to be feasible (HHS 

2005).

For any claim referred to NIOSH regarding an employee, 

• who was employed during the Cohort period but because of limited employment during 

this period is not a member of the Cohort, or

• who is a member of the Cohort and whose cancer is not defined as a specified cancer 

under EEOICPA (and so is not eligible for compensation under EEOICPA without a 

dose reconstruction), 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 11 of 102

NIOSH will continue to attempt to complete a dose reconstruction for the exposure period based 

solely on external and medical X-ray radiation sources. However, because of the SEC determination 

(HHS 2005) that it is infeasible to adequately reconstruct internal dose during the period October 1, 

1942 through October 31, 1947, dose estimates for this period are considered partial dose estimates.

1.2 SCOPE

DOL has designated the Linde Ceramics Plant a DOE facility and the Tonawanda Laboratory an AWE 

facility (Turcic 2007). This document covers both facilities. The information in this site profile

supports the assumed operational and residual contamination periods listed below. DOL has 

determined that the residual contamination period for the Tonawanda Laboratory is also applicable to 

the Ceramics Plant (Turcic 2008). Although cleanup activities at the Ceramics Plant continued into 

July of 1954, the designated covered period for this facility ends in 1953. Post-1953 exposures are

also covered under the EEOICPA, but this period is termed the residual exposure period. Because

the activities and exposure potential at the Ceramics plant during the first part of 1954 (January 1 

through July 7) are the same as in the immediately previous period (1950 to 1953), information on 

reconstruction of dose for the period from January 1 through July 7 is included in the operational 

period section of this document. The instructions in this document for reconstruction of dose at the 

Ceramics plant during the residual period (as defined by DOL as starting on January 1, 1954) pertain 

to exposures starting after July 7, 1954. July 7, 1954 is used as the definitive end of the 

decontamination period at the Ceramics plant based on the date of the final survey of the facility, 

which is documented in a memorandum from the New York Operations Office (NYOO) to Union 

Carbide that asserts that the decontamination requirements of the contract were fulfilled (Eisenbud 

1954).

Facility Operational period Residual contamination perioda

Linde Ceramics Plant and 

Tonawanda Laboratory

10/01/1942–10/31/1953

1988–1992; 1996 (remediation)

January 1, 1954 to July 2006b

a. Also called the postcleanup period in this document.

b. Excluding 1988 to 1992 and 1996.

Section 2.0 describes the site and its operational history. Sections 3.0 and 4.0 describe estimation of 

internal and external exposure from 1942 to July 7, 1954, respectively. Section 5.0 describes 

occupational medical exposure. Section 6.0 provides information on exposures during the residual 

contamination period after 1953. Attributions and annotations, indicated by bracketed callouts and 

used to identify the source, justification, or clarification of the associated information, are presented in 

Section 7.0.

Attachment A contains data that was used in analyzing exposures of workers to beta radiation. 

Attachment B lists codes and special terminology in the LAPC records. Attachment C shows data 

sources on uranium progeny concentrations, and Attachment D provides a uranium coworker 

assessment for November 1947 to January 1950. Attachment E provides an assessment of dose 

consequences from uranium ore bag that were stored on the site during the postoperations period.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 12 of 102

2.0

When the U.S. government and its contractors first became interested in uranium, LAPC, then a 

division of Union Carbide and Carbon Corporation, operated Tonawanda Laboratory, which had been 

producing U3O8 that was sold as a coloring agent for ceramics. Because of the great interest in 

obtaining uranium that could be used to create experimental uranium piles, Linde was contracted to 

develop uranium chemical processes and build a facility that could process large amounts of uranium 

ore. This commissioned facility was called the Linde Ceramics Plant. The plant worked with two 

types of radioactive material:

SITE DESCRIPTION AND OPERATIONAL HISTORY

• Refined uranium materials and preprocessed (partially refined) domestic ores, and

• Unprocessed African ores.

Processing of the African ores involved exposures to significantly higher levels of 230Th, 226Ra, and 

222Rn. In the other materials, the initial refining process probably removed most nonuranium 

radionuclides. Due to the long half-lives of 230Th and 231Pa, significant ingrowth of these nuclides and 

their progeny in the 238U and 235

U decay chains did not occur in the interval between the refining of the 

materials and their processing at Linde.

LAPC performed work for the Manhattan Engineer District (MED) and its successor the U.S. Atomic 

Energy Commission (AEC) between 1942 and 1949. Tonawanda Laboratory performed research and 

development (R&D) on uranium processing for the MED beginning in late 1942 and probably ending 

in 1946. In the early years, it operated pilot plants to develop procedures for the Ceramics Plant. The 

Ceramics Plant produced uranium materials for the MED and AEC from 1943 to 1946 and from 1947 

to 1949. From 1947 to 1949 (and perhaps earlier), Linde received UO2 for processing from 

Mallinckrodt Chemical Works (AEC 1949a). The plant also produced nickel material for the K-25 

diffusion barrier. The Ceramics Plant was in standby from mid-1946 to late 1947. The end of 

production in 1949 was followed by cleanup and decontamination and then turnover of the production 

facilities back to Linde. This turnover probably occurred in 1954. 

Uranium processing at the Ceramics Plant involved three steps: 

• Step I, conversion of ore to U3O8 (black oxide);

• Step II, conversion of U3O8 to UO3 (orange oxide) as an intermediary product and then to UO2

(brown oxide);

• Step III, conversion of UO2 to UF4 (green salt).

The following sections provide more detailed histories of Linde facilities, radiation sources, and 

processes that related to MED/AEC radiation exposures. Key dates are noted in Table 2-1.

2.1 FACILITIES

The Linde Ceramics Plant and Tonawanda Laboratory were on land that was owned by Union 

Carbide at East Park Drive and Woodward Avenue in Tonawanda, New York (ORNL 1978, Fig. 3). 

The site is near the intersection of Riverview Boulevard and Woodward Avenue. It is north of 

Woodward Avenue, east of East Park Drive, and west of the Conrail railroad tracks.
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Table 2-1. Key dates.

Assumed date or perioda Event or activity

Ceramics Plant

06/01/1943–07/31/1946b Step I production (U3O8 from ore or sludge)

06/01/1943–11/30/1943 Preprocessed ores (domestic and scrap)

12/01/1943–11/12/1944 African ores

11/13/1944–01/31/1946 Preprocessed domestic ores

02/01/1946–02/28/1946 African ores (48%) and preprocessed domestic ores (52%)

03/01/1946–06/30/1946 African ores

07/01/1946–07/31/1946 African ores and preprocessed ash

04/27/1943–03/08/1944c Step II production (UO2 from U3O8)

07/25/1943–06/26/1946d Step III production (UF4 from UO2)

08/01/1946–09/14/1947 Standby

09/15/1947–10/31/1947e Step III rehabilitation

11/01/1947–06/30/1949f Step III production

10/01/1944–02/28/1946g Production of nickel material for K-25 diffusion barrier

07/01/1949–03/31/1950h Cleanup of Building 30

07/01/1949–12/31/1954i Cleanup of Linde Ceramics Buildings

Before 1978 Demolition of Buildings A and B (office buildings on Figure 2-1)

After 1954 Postcleanup period

08/31/1981j Demolition of Building 37 began

08/31/1996j Demolition of Building 38 began

09/05/1998k Demolition of Building 30 began

09/30/2000j Soil remediation began (scheduled for completion in 2007)j

Tonawanda Laboratory

10/01/1942–07/31/1946l MED-related R&D operations

08/01/1946–12/31/1946m Cleanup period

After 1947 Postcleanup period

04/30/2004j Demolition of Building 14 began

a. Unless more precise information was available, activities were assumed to begin on the first day of the start 

month and to end on the last day of the completion month in the data source, and demolition activities were 

assumed to begin on the last day of the start month in the data source.

b. All Step I dates and data are based on Table B-1 of Aerospace Corporation (1981) except for the November 

1944 transition date from African to domestic ore. The date was determined from analysis of Step I film badge 

data (see Section 4.1.3.2).

c. LAPC (1946a, p. 40).

d. LAPC (1946b, p. 42).

e. Start date based on Kent (1947, p. 1).

f. Start date based on Rennich (1947). End date based on Heatherton, 1948f.

g. Hickey, Crawford-Brown, and Tankersley (1988, Figure 13, p. 2).

h. Start date assumed to be day after end of Step III production; end date estimated based on Eisenbud (1950).

i. Start date assumed to be day after end of Step III production; end date estimated based on Harris (1954).

j. Pilon (2004).

k. USACE (1998).

l. Start date based on Jenness and Ewing (1943). End date assumed to be the end date of 1943 to 1946 

production at the Ceramics Plant.

m. Start date based on estimated end date of R&D period; end date estimated.

Tonawanda site buildings that were involved in MED/AEC work are shown in Figure 2-1. The 

Tonawanda Laboratory occupied Building 14, and the Ceramics Plant used Buildings 30, 31, 37, 38, 

and A. Building B contained MED offices. Ownership of the Ceramics Plant buildings was transferred 

to LAPC after the site cleanup that began with the shutdown of production in 1949. The transfer 

probably was completed in 1954. In the 1990s, the site was acquired by Praxair (USACE 2003). 
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Figure 2-1. Linde Ceramics Plant and Tonawanda 

Laboratory buildings on LAPC Tonawanda site. 

2.1.1

Building A was the Linde Ceramics Plant office building for administrative and support personnel. 

During the MED years, Building B housed the Tonawanda Area Engineer office of the MED (Dupree 

1983a) and probably was later used by AEC personnel. The designations A and B were adopted for 

this document and might be different from the official Linde or MED designations. Buildings A and B 

were no longer standing in 1978, when the results of a 1976 radiological survey were issued by Oak 

Ridge National Laboratory (ORNL 1978). 

Buildings A and B

Figure 2-1 is based on Figures 3-1 and B-11 of BNI (1982), Figure 3 of Frame et al. (1981), and LAPC

(1945a).

A shelter in which workers were allowed to smoke was at the south end of Building B (Dupree 1983a). 

This was probably for use only by office workers. A separate smoking shelter was provided for 

production workers (Klevin 1949a, data sheets 543 and 546).
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2.1.2

Building 14 housed Tonawanda Laboratory. The U.S. Army Corps of Engineers (USACE) has 

described its use as follows (ACE Buffalo, 2003):

Building 14 (Tonawanda Laboratory)

Building 14 was used for laboratory and pilot plant studies for uranium separation in the 

early part of MED operations. Historical drawings indicate that the MED laboratory and 

pilot plant studies were initially confined to the south part of the building. It is unclear 

how extensively the remainder of the building was used for MED operations. However, 

documents indicate that laboratory and pilot plant operations were continued for the 

purpose of experimenting and developing more efficient processing methods, and 

operations appear to have been expanded into most of the building, possibly to support 

larger pilot studies. The available records do not indicate whether or not the use of 

Building 14 ceased before the MED/Atomic Energy Commission (AEC) operations 

were discontinued at Linde.

A 1942 Tonawanda Laboratory directory listed approximately 200 employees (LAPC 1942). The 

locations of most were designated by room numbers. Some employees were said to be in the Eng. 

Lab and others in the Proving Lab. A 1944 employee list reported some to be in the Pilot Plant (LAPC

1944a). The terms Proving Lab and Pilot Plant appear to refer to the same facility.

Tonawanda Laboratory is also referred to by other names such as Linde Research Laboratory 

(Dupree 1983a) and The Laboratory of The Linde Air Products Company (Jenness and Ewing 1943).

Contamination in Building 14 was found in a 1976 radiological survey (ORNL 1978). At the time, 

Building 14 housed fabrication facilities, research facilities, and offices generally used by 20 to 30 

employees. Demolition of Building 14 began in April 2004 (Pilon 2004). 

2.1.3

Building 30—also called the uranium refinery—was used for Steps I and II. Figures 2-2 and 2-3 show 

the layout. It had two floors. Large pieces of equipment such as digestion tanks were on the ground 

floor. Process workers operated equipment from the second floor (Dupree 1983a,b). The 1949 to 

1950 decontamination work by LAPC and the results are described in Heatherton (1950). As of 1976, 

the building was used as a shipping and receiving warehouse and occupied by about 20 to 30 

employees. Radioactive contamination was found during a 1976 survey (ORNL 1978). Building 30 

was demolished during the period from September 5 to 19, 1998 (USACE 1998).

Building 30 (Ceramics Plant)

2.1.4

Building 31 is described as "used in uranium separation process during MED operations" (BNI 1993, 

p. 1-65). A partially readable 1945 plant drawing indicates that it contained shipping and storage 

areas, maintenance facilities (including a welding shop, a machine shop, a gauge test room, and a 

carpentry area), and a small number of associated offices (LAPC 1945a). Building 31 might have 

been the final location of the Linde Tonawanda nickel processing operations.

Building 31 (Ceramics Plant)

As of 1976, the building was described as housing fabrication facilities, offices, and storage areas, 

and it was occupied by about 12 to 15 employees. Radioactive contamination was found during a 

1976 survey (ORNL 1978). Building 31 was decontaminated in 1997 (USACE 2002). As of 2004, it 

was still standing and in use (Pilon 2004).
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Figure 2-2. Building 30 layout (Heatherton 1950).

2.1.5

Building 37 was a small appendage to Building 39, which was east of Building 38 (Figure 2-1). 

According to BNI 1993 (p. 1-65), Building 37 was used for Step III. However, no details of this use 

have been found and its small size, approximately 16 by 36 ft (BNI 1982 Figure B-11), indicates at 

most a minor role. The 1945 plant drawing labeled it Test Bldg (LAPC 1945a). Radioactive 

contamination was found in 1976 (ORNL 1978). Linde demolished Building 37 in 1981 (Pilon 2004).

Building 37 (Ceramics Plant)

2.1.6

Building 38 was used for Step III. Figure 2-4 shows its layout. Radioactive contamination was found 

in 1976 (ORNL 1978). Building 38 was demolished in 1996 (Pilon 2004).

Building 38 (Ceramics Plant)

2.2 SOURCE TERMS FOR AFRICAN AND DOMESTIC URANIUM ORES

Linde used two types of starting materials for Step I (ore to U3O8) processing. These are generally 

referred to as African ore and domestic ore in Linde-related literature. Most of the African ore was 

pitchblende; some was torbernite (Aerospace Corporation 1981, p. 1). Neither African ore underwent 

chemical preprocessing before being shipped to Linde. Therefore, they contained not only uranium 

but also all members of the uranium decay series including radium and its progeny.

The domestic ore category included preprocessed domestic ores and byproducts of other uranium 

conversion processes. The preprocessed domestic ores derived from tailings from vanadium 

processing. The preprocessing concentrated the uranium and had the effect of removing "the major 
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Figure 2-3. Building 30 layout (LAPC undated a).
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Figure 2-4. Building 38 layout (Klevin 1949a).

portion of the radium" in the original ore (Aerospace Corporation 1981, p. 5). The preprocessing 

might have also separated uranium from other radionuclides in the ore. If the separation of uranium 

had been complete, then the radioactivity in the preprocessed ore delivered to Linde would have 

consisted primarily of 234U, 235U, and 238U (in their natural abundances) and their short-lived progeny 

that had grown in after the extraction process. The completeness of the separation of uranium from 

other radioactivity in the vanadium tailings Linde used is not known. A review of the chemical 

processing concluded only that "the radium in the domestic ore was significantly less than that in the 

African ores" (Aerospace Corporation 1981, p. B-1).

The distinction between the two types of ore is important. The primary radiological hazards from an 

ore that contained only uranium and its short-lived progeny would be due to alpha and beta 

emissions. In contrast, radium and other progeny in the African ores would produce, in addition to the 

alpha and beta emissions, significant gamma emissions and elevated levels of radon. In this 

document, the term "domestic ore" always means preprocessed domestic ore, and is sometimes 

referred to as "refined ore." "African ore" always means ore that has not undergone preprocessing.

Enriched uranium (up to 35% or less by mass) might have been used on a limited basis for K-25 

barrier testing and R&D at Tonawanda Laboratory. Because of the very limited quantities in 

comparison to the ore, only natural-abundance uranium is explicitly considered for this site. 

2.3 ACTIVITIES

LAPC is reported to have begun research into the processing of uranium ore in the late 1930s before 

it had a contractual relationship with the MED. At the time, uranium was used to make colored glazes 
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for ceramic dinnerware. A former MED employee interviewed in 1983 (Dupree 1983a) said that Linde 

Research Laboratory in Tonawanda, New York, developed a process to obtain purified U3O8 from 

carnotite ore and produced 80 t of U3O8. Linde's capability to provide commercial quantities of U3O8

as of 1941 was confirmed in a letter to atomic weapons physicist Leo Szilard (Dana 1941).

On November 16, 1942, LAPC entered into contract W-7401 eng-14 with the MED to design and 

construct facilities at the Ceramics Plant for processing uranium ores and to produce U3O8, UO2, and 

UF4 (Cornell 1942; Marshall 1943). The contract also called for R&D into processes that would be 

used in the Ceramics Plant facilities.

2.3.1

Available reports indicated two main areas of Tonawanda Laboratory MED uranium work. The first 

was development of methodologies for Ceramics Plant operations. A report on the first year of work 

covered the period from October 1942 to November 1943 (Jenness and Ewing 1943). Most of the 

work was devoted to study of the problems in the design, operation, and improvement of the 

Ceramics Plant.

Uranium Research and Development

Three different large-scale pilot plant programs were described. As of November 20, 1942, one was 

producing 2 to 3 t of U3O8 per week (Bonsib 1942). The third program, which involved African ore, 

had begun in October 1943 and was expected to conclude in December 1943.

Available documentation from the Laboratory includes three research reports (Brimm 1943a; Skinner 

1944; Wiesendanger 1944) and a final report (LAPC 1946c). The final report discussed work on 

R-10, a low-grade African ore, and Q-20, a natural (i.e., not preprocessed) torbernite ore. No dates 

for this work were provided, but because R-10 and Q-20 processing did not begin at the Ceramics 

Plant until February 1946, the R-10 and Q-20 studies at Tonawanda Laboratory probably were 

performed in late 1945.

The second area of Tonawanda Laboratory work related to conversion of UF6 process gas to UO3. 

Two research reports (Brimm and Schubert 1945a,b) and a final report (LAPC 1946d) indicated that 

the associated pilot plant operations used pounds rather than tons of material.

A later Linde research report on uranium processing from September 29, 1948 was prepared by 

Ceramics Plant rather than Tonawanda Laboratory personnel (Chapman et al 1948). The absence of 

MED or AEC reports after May 1946 suggests that MED work by Tonawanda Laboratory ended then 

and that there was no later AEC work. For dose reconstruction, it is assumed that there was no 

Tonawanda Laboratory work for the MED or AEC after July 31, 1946, the end date of the initial 

production period at the Ceramics Plant.

2.3.2

The Linde Ceramics Plant engaged in three different uranium production activities — Steps I, II, and 

III. Step I (production of U3O8) typically involved ores that contained 3% to 20% U3O8 by weight. 

Approximately 70% of the ore was African and 30% domestic by weight. About 26,000 MT of ore 

were processed, and about 2,300 MT of U3O8 were produced (Aerospace Corporation 1981, 

Table B-1). 

Uranium Production

Step I (production of U3O8) took place in Building 30 from June 1943 through July 1946. Step II 

(production of UO2) took place in Building 30 from April 1943 through March 1944. Step III 

(production of UF4) took place in Building 38 from July 1943 to June 1946. Linde sent the U3O8 from 

Step I to DuPont and Mallinckrodt after Step II production ceased (Gates 1946). Linde received UO2

from other companies for Step III processing (AEC 1949a). Tonawanda Laboratory might have 
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investigated all of these processing steps in its pilot plant. Uranium production was in standby after 

July 1946. Preparations for Step III resumption began in September 1945. Step III operations were 

resumed under contract AT-30-1-GEN-165 in November 1947 (Rennich 1947). Linde ended uranium 

production for the AEC in June 1949.

A MED record-of-negotiations memorandum (Dreveskracht 1945) noted that Linde had offered for 

sale 36,000 lb of uranium ore concentrates from Colorado and 8,900 lb of residue, which were the 

"property of Contractor as a result of operations before his present CPFF [cost plus fixed-fee] work for 

the MED (Contract No. W-7401 eng-14)." The transfer was made under contract W-17-028 eng-29 on 

May 4, 1945 (USEO 1945). Correspondence indicates that Linde transferred ownership of the 

uranium concentrates and residues to the government on August 1, 1945 (Martin 1945), and the 

material was subsequently processed in the Ceramics Plant for MED (Dupree 1983a, Robinson 

1945).

Detailed descriptions of the typical Ceramics Plant uranium processing methods are presented below. 

These include descriptions of chemical and operational processes.

2.3.2.1 Step I (Uranium Ore to U3O8)

Step I extracted purified U3O8 (black oxide) from uranium ore. The Step I process varied with time as 

the nature of the ore changed and as Linde gained experience with the procedure. 

Domestic ore was ground and mixed with water to form a slurry. Sulfuric acid and other chemicals 

were added, and the mixture was digested at 90°C for 2 to 3 hours and then cooled to 60°C. This left 

the uranium in solution as uranyl sulfate along with some of the impurities. Soda ash and sodium 

bicarbonate were added to make the solution basic. This precipitated most of the remaining 

impurities and left the uranium in solution as sodium uranyl tricarbonate. The slurry was filtered in 

Moore filters (described in the next section). The liquors contained the uranium and some 

objectionable impurities. Vanadium and phosphorous impurities were precipitated by the addition of 

ferrous and ferric sulfates and then removed in a second filtering in plate and frame presses, which 

produced “iron cake.” The liquors were treated with caustic soda to precipitate the uranium as sodium 

diuranate, which was removed as cake in a third filtering process. The sodium diuranate cake was 

treated with sulfuric acid and ammonium sulfate to convert it to an ammonium uranyl sulfate complex. 

This was removed in a fourth filtering process. The resulting “acid leach cake” was fed to a calciner to 

drive out the ammonia, sulfur dioxide and trioxide, and water to leave the black oxide (U3O8).

A modification of the above procedure was used to treat African ore. The filter cake from the Moore 

filters (first precipitation step) contained the radium in the ore. In the second step, instead of iron 

sulfates, sodium sulfide was added to remove lead. Some of the other chemical additives that were 

used at various points were different (LAPC 1946e; Aerospace 1981, p. A-1).

Step I Operations

Ore was delivered by rail in boxcars. African L-30 pitchblende arrived in burlap bags inside paper 

bags. Each ore bag weighed 50 to 80 lb. For African ores, the car doors were left open for 12 to 

24 hours before unloading to reduce radon concentrations in the car. Either ore bags or bulk ore (the 

practice varied) were manually loaded onto wheeled carts (buggies) by workers called loaders or 

movemen and transported to a storage or processing area (Olevitch 1944; LAPC 1946e; Cranch 

1944a).

For dry ore, processing began in the dumping room on the receiving platform. Ore was dumped 

through a grate onto a conveyer belt that carried it to the ore storage bin. The storage bin held about 

75 t of ore. Ore from the bin was fed to a ball mill by a Feedoweight (a belt conveyor that weighed the 

ore). After milling, the ore was fed to the digestion process.
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Wet ore was transported to the second floor using the freight elevator and either dumped into the 

scoop box of the ball mill or directly into digest tanks (skipping the ball mill step). Loaders did not ride 

on the elevator with the ore.

After the dumping, about a 0.5 lb of ore remained in each bag. To recover it, the bags were shaken in 

a bag shaker over a conveyor belt onto which freed ore dropped. The bags were then washed, 

centrifuged, and burned in an incinerator. The washed bags contained about 10% to 30% of the 

remnant ore. At times, washed bags that were awaiting incineration were stored outdoors on the site. 

A July 1944 report noted that 19,000 bags were currently being stored (Olevitch 1944, Appendix A).

For digestion, chemicals appropriate to the ore type were added to the digest tanks, and the mixture 

was heated and cooled to precipitate components. Precipitates were separated from solutions using 

various filter assemblies. The first filtration in the Step I processing used the Moore filters. A Moore 

filter consisted of a large basket that contained 24 rectangular leaves with each leaf approximately 35

ft2 in area. Each leaf was formed by a rectangular perforated pipe frame over which a bag was 

stretched. The filter basket was placed by crane into a neutralized digest tank and held there while 

suction was applied to the pipe. Gelatinous cakes about 1-in. thick formed on the leaves. The basket 

was removed from the digest tank, placed in a wash tank, washed, and then moved to a hopper. The 

suction was then reversed to inflate the bags and cause the cake to fall into the hopper. From there 

the cake was transferred to an open-top truck below the hopper for transport to a storage location. 

After sampling, the filter cake was hauled away by truck. Radium-containing filter cake from the 

African ores was shipped to the Lake Ontario Ordnance Works (LOOW) in Lewiston, New York, for 

storage. Filter cake from the domestic ores went to a tailings pile. From some filter presses, the filter 

cake was manually dumped by the operators into buggies for transport to the next step. From other 

presses, the cake could be fed automatically to the next step (MED undated, page 3; Olevitch 1944; 

LAPC 1946e; Aerospace 1981, p. A-1).

The tanks used in the process were large: 18 to 31 ft in diameter and one story high (Dupree 

1983a,b).

2.3.2.2 Step II (U3O8 to UO2)

Step II converted U3O8 to UO2 (brown oxide) with UO3 (orange oxide) as an intermediary product. 

Black oxide (U3O8) was dissolved in nitric acid. Some insoluble impurities were filtered off, and the 

solution was evaporated to molten uranyl nitrate hexahydrate. The next step was based on ether 

extraction of the uranium. The extraction took advantage of the fact that metallic salts are more 

soluble in water than in ether, but the uranyl nitrate is nearly equally soluble in either and can be 

shifted from one medium to the other by changes in volume and concentration ratios. The uranyl 

nitrate hexahydrate was added to cold ether, and the solution was washed with water. The impurities 

went to the water layer and the uranyl nitrate to the ether layer. The purified uranyl nitrate was 

denitrated to powdered uranium trioxide (UO3) by heating in a stainless-steel pot furnace. The trioxide 

powder was collected mechanically and flowed by gravity through a micropulverizer into a rotary tube 

reduction furnace. There the powder was reduced to the desired uranium dioxide (UO2) using 

gaseous hydrogen as a reducing agent. The finished UO2 product was packaged, sampled, sealed, 

and stored (LAPC 1946a; Aerospace 1981, p. A-1).

Step II Operations

U3O8 in buggies from Step I was weighed and hoisted to the second floor. It was then fed into nitric 

acid in digest tanks with scoops. The digest tanks were heated to 90°C. After digestion, insoluble 

impurities were filtered out using a filter press. Liquids were piped from one vessel to another. 

Frequent chemical analyses of samples were required in this process to determine the progress of 
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chemical reactions. The processing produced various liquors and some associated cakes that were 

designated "OK" or "NG" (LAPC 1946a).

2.3.2.3 Step III (UO2 to UF4)

Step III converted UO2 to UF4 (green salt) using the chemical reaction

UO2 + 4HF  UF4 + 2H2O

The reaction was carried out at 1,000°F. Gaseous anhydrous HF gas was flowed over the UO2, 

which was placed on magnesium trays stacked on spacer bars inside 9-in.-square tubes 10 ft long 

that were inside a furnace (LAPC 1946b; Hickey, Crawford-Brown, and Tankersley 1988, p. 8).

Step III Operations

UO2 was weighed and hand-troweled into shallow trays inside a hood with a dust collector operating. 

Twelve pounds of UO2 were placed on each tray. The oxide was furrowed when loaded to maximize 

surface area. The trays were transported on buggies and inserted into the furnace. 

After loading, the furnace was sealed, purged of air, and heated to the starting temperature. Then the 

flow of HF gas was started. After the required number of hours, the heat and gas flow were stopped, 

and the furnace was purged of HF gas, cooled, and opened.

The trays were removed from the furnace and hand-placed onto buggies for transport to and 

placement inside an unloading hood with a dust collector. Good material (light green) was loaded into 

hoppers. Bad material (dark green) was placed in fiber-pack drums for later retreatment. Good 

material was pulverized, blended, sampled, packaged, weighed, and shipped (LAPC 1946b; Hickey, 

Crawford-Brown, and Tankersley 1988, p. 8).

2.3.3

After the shutdown of Step II in 1944, the Step II equipment was converted for use in production of 

nickel powder, a nonradioactive process, which occurred from October 1944 through February 1946 

(Hickey, Crawford-Brown and Tankersley 1988, pp. 9 and 13). The nickel was for use in the K-25 

diffusion barrier (Hunter 1949; Dupree 1983b, p. 3). 

Nickel Production

There is discrepant evidence on the location of the nickel processing work. A contractual record (USA

1945) states that Linde was contracted on April 7, 1944, via letter contract W-7418 eng-51 to 

construct and equip a plant to produce 80 t of special nickel oxide per month and that the plant was to 

be constructed in Tonawanda on the same premises and adjacent to the buildings that were used in 

connection with contract W-7401 eng-14 (the contract for the uranium refining work). A drawing that 

was attached to a former Tonawanda Area Engineer Office employee interview record (Dupree 

1983a) shows the location of the nickel processing to be in a building east of Building 30. Therefore, 

nickel processing might have occurred in Building 31 (see Figure 2-1). However, another record of 

interview (Dupree 1983b) states that when Step II operations in Building 30 ended, the Step II area 

was used to refine nickel. Both recollections could be correct. It could be that nickel processing 

initially occurred in the Step II area and later was moved to Building 31.

Evidence that supports Building 31 as at least the ultimate location for Linde's Tonawanda nickel 

processing is found in a 1945 plant drawing (LAPC 1945a). The visible portions of a partially 

obscured label on the drawing are consistent with the interpretation that it says "Nickel Plant." In 

addition, the drawing shows an ammonia cracking area in Building 31, and other records indicate that 

ammonia was used in the nickel processing [chemical operators in department C2P used ammonia 
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according to LAPC (1945b, p. 17), and department C2P engaged in nickel processing according to 

Hickey, Crawford-Brown, and Tankersley (1988, p. 28)].

Nickel processing at the Ceramics Plant has been described as follows (Hickey, Crawford-Brown, and 

Tankersley 1988, p. 9):

Metallic Ni slugs (small ingots) were oxidized with HN03, filtered, and heated to 1200 C 

to produce a nickel oxide (NiO). This was pulverized, mixed with ammonium chloride 

(NH4Cl), and heated to form a porous mass of the oxide, which was reduced to 

powdered nickel by exposure to H2. The powder was pulverized and packaged. ... It is 

not known whether both NiO and powdered Ni or only NiO were produced in quantity at 

this plant.

2.3.4

Early on, radioactive liquid wastes were discharged to the Tonawanda sanitary sewage system. Due 

to the nature of the liquids, this became a problem, and Linde began to dispose of liquid wastes in 

onsite wells that sometimes overflowed. Later still, liquid wastes were discharged to a drainage ditch 

that led to a sewer conduit (BNI 1993, pp. 1-9 to 1-15). 

Other Radiological Activities

After MED work began at the Tonawanda site, there was potential exposure of workers to radiation 

and radioactivity when outdoors. Portions of the site are known to have been contaminated with 

radioactivity (Heatherton 1948a; ORNL 1978; BNI 1982); resuspension would have produced airborne 

radioactivity. One source of ground contamination and airborne radioactivity was the ore unloading 

process, which involved transporting ore in buggies – sometimes in bulk and sometimes in bags –

from boxcars to Building 30 (see Step I Operations in Section 2.3.2.1). Outdoor areas of the site were 

sometimes used for storage of radioactive materials. Olevitch (1944) reports the outdoor storage of 

contaminated ore bags that at times numbered in the thousands. In 1948, 1 mR/hr gamma and 

3,000 α dpm/100 cm2 were measured from the soil in an area formerly used for storage of radioactive 

materials (Heatherton 1948a).

An additional source was the release of liquid effluents either to onsite wells that sometimes 

overflowed, or to an onsite drainage ditch (BNI 1993, pp. 1-9 to 1-15). Airborne effluents from the 

plant were an additional source of outdoor radioactivity.

In 1949, Linde workers unloaded drums of K-65 product that was shipped by rail from Mallinckrodt in 

St. Louis to the LOOW at Modeltown, New York (Wolf 1949b; Heatherton 1949a; NYOO 1949). Linde 

film badges were worn during this work.

At the April 18, 2005, Worker Outreach Meeting, strontium was mentioned. At the June 27, 2005,

Worker Outreach Meeting, it was clarified that the strontium was in the form of a 90Sr sealed source. 

This source was reportedly brought to the site in the late 1950s. The estimates of uranium exposure 

based on estimates of exposure periods and source term, which were based on worst-case 

assumptions when a parameter was not well supported by available information, would be sufficiently 

bounding to account for small amounts of radioactive strontium.

2.4 PROTECTIVE MEASURES

Steps were taken to protect against radiological and chemical hazards during processing (LAPC

1945, 1946a,b,e; Dupree 1983a,b). Although some protective measures were included early in the 

Linde program, protective measures were upgraded over time in response to both survey findings and 

newly available information about biological risks of exposure. These measures included limiting the 

time a worker was allowed to spend in a radiologically hazardous activity, provision and upgrading of 
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ventilation to reduce concentrations of radon and radioactive dusts in the air, and good housekeeping 

to reduce radioactive dust on the floor. Respirators to reduce dust and chemical exposure and gloves 

to reduce beta exposure to the hands were required for some of the more hazardous tasks, but these 

safety devices were not used in all process areas by all workers. Some workers changed into 

company-issued clothing when they arrived at work and changed back to their own clothing at the end 

of a shift. Showers were available. At some point in Linde operations, showering at the end of a shift 

became mandatory for some workers, and they were "wanded" by a health physicist to check for 

radioactivity when they left the shower.

Chemical protective measures included goggles, respirators, rubber gloves, face shields, rubber or 

plastic-coated aprons and sleeves, and head caps. There were more substantial protective measures 

for operators who worked with hydrofluoric acid. 

According to interviewees, use of respirators was often at the worker's discretion, but they were 

"usually used" (Dupree 1983b). Inspectors reported mixed compliance with recommendations to wear 

respirators. The author of a June 1944 inspection report commented, "The wearing of respirators 

remains difficult to enforce" (Ferry 1944a). The author of an October 1944 inspection report 

observed, "Respirator discipline was about 80% efficient in the dumping room" (Tybout 1944a). In 

1945, the same inspector witnessed the collection of dust samples by Mr. H. Seemann, Assistant to 

the Safety Engineer. In his report, the inspector noted, "The men wore respirators, but Mr. Seemann 

stated that this was the case only because he was present" (Tybout 1945a). A 1948 dust sample 

collector commented that when samples were collected, individuals on the day shift wore respirators 

but those on the evening shift did not (Hayden 1948).

No credit for safety gear (including respirators) is taken in this document.

2.5 PERSONNEL, JOB CATEGORIES, AND WORK HOURS

For Tonawanda Laboratory, two listings of employees might aid in determining whether a claimant 

worked there. A September 1942 Laboratory directory (LAPC 1942) lists approximately 200 

employees. Several employees are linked to the Proving Lab, which might be an indicator of higher 

exposures once the uranium work began. A 1944 employee list (LAPC 1944a) contains 

approximately 120 names and similar information. It is not clear whether this list includes all 

employees.

For the Ceramics Plant, an April 1944 employee list contains nearly 400 names (Neuman 1944) and 

specifies the workers’ jobs. A December 1945 job description specifies the duties of 51 categories of 

workers and lists a department code for each position (LAPC 1945b). Because the same title (e.g., 

chemical operator) was sometimes used in different departments in which the nature of the work was 

very different (e.g., Step I and nickel processing), knowing the department might help identify the type 

of activity in which a worker was involved. Table 2-2 shows Ceramics Plant department codes that 

were associated with various worker activities as determined in a 1988 study of worker hazards

(Hickey, Crawford-Brown, Tankersley, 1988). A 1949 study reported that there were 139 Step III 

personnel and listed the number in each job. Names of some of the personnel were provided (Klevin 

1949a, pp. 6, 80). 

Table 2-2. Department codes for various worker activities.

Department Department codea

Administrative services AB, AD, AL, AM, BA, BB, BT, A/M

Chemical control, R&D BL, HL

Engineering and development AW

Maintenance AN

Process Operations: Step I CL
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 25 of 102

Step II CM

Step III CN

Nickel CP, CR, C2M, C2P

Loading: General AN, AH, LOOW

Janitors AA, GE

Stores and supplies AS, ASA, A8A, AT

Safety and security AJ, A9A, A2J

Undetermined activity CA, CS

Untitled jobs and departments (b)

a. From Hickey, Crawford-Brown, and Tankersley (1988, p. 27). Based in part on a 

plant document identifying some of the codes and in part on inference from less 

direct information in plant records.

b. Not available.

The term “day” in this document refers to a calendar day. The term “workday” (wd) is used to 

describe a day at work. Default assumptions are 250 wd/yr and 8 hr/wd: this results in 2,000 hours of 

work per year. The distinction between workday and calendar day is especially important when 

considering internal dose rates for use in calculating organ doses, because intakes actually occur 

during the workday, but dosimetry calculations are usually based on integration over calendar days.

The Ceramics Plant work schedule during the production period (1943 to 1946) is described as 

involving 8-hour shifts 6 d/wk (Dupree 1983a,b,c). A 9-hour workday with a lunch period included is 

assumed because there is evidence that many employees worked overtime (Dupree 1983b, p. 4; 

MED undated, p.1). An 8-hour workday plus a half-hour for lunch is known to have been in effect for 

many of the workers in late 1948 (Klevin 1949a). The transition from a 6-day workweek to a 5-day 

workweek was assumed to have occurred on January 1, 1951, based on a report by interviewees that 

the 48-hour workweek lasted until 1950 (Dupree 1983b, p. 4).

Table 2-3 provides a list of some Linde references that include worker names and sometimes 

additional job information.

Table 2-3. Worker information.

Item Reference

Tonawanda Laboratory employees and room numbers, September 2, 1942. Job 

titles are not stated.

LAPC 1942

Linde employees associated with contract W-7401-eng-14, July 20, 1943. Some 

are Tonawanda Laboratory personnel. Some might be Ceramics Plant personnel. 

Job titles are included.

Pew 1943

Ceramics Plant personnel with job titles, April 25, 1944. Neuman 1944

Ceramics Plant Step III process operators, June 14, 1948. Heatherton 1948b

Ceramics Plant Step III process area personnel (including operators, millwrights, 

and stores attendants), January 25, 1949.

Klevin 1949a, p. 50

Film badge exposure report listing 23 Step III personnel by name and job title, for 

week beginning March 28, 1949.

AEC 1949b

List of Ceramics Plant Step III process operators, August 26, 1949. Klevin 1949b, p. 11

Based on the above, the periods and work schedule at the Ceramics Plant from the onset of MED 

operations through July 7, 1954, were assumed to have been as shown in Table 2-4.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 26 of 102

Table 2-4. Ceramics Plant assumed work schedule (including lunch and 

breaks), 1942 to July 7, 1954.

Period Start End hr/wd wd/wk wk/yr

Preproduction 10/01/1942 04/26/1943 9.0 6 50

Production 04/27/1943 07/31/1946 9.0 6 50

Standby 08/01/1946 09/14/1947 8.5 6 50

Rehabilitation and production 09/15/1947 06/30/1949 8.5 6 50

Cleanup 07/01/1949 12/31/1950 8.5 6 50

Cleanupa 01/01/1951 07/07/1954 8.5 5 50

a. Assumed date of transition from 6- to 5-day workweek (based on Dupree 1983b, p. 4).

2.6 DECONTAMINATION DURING THE MED/AEC CONTRACT PERIOD

Cleanup and decontamination activities at the Ceramics Plant facilities during the MED/AEC contract 

period are discussed below. No information was found on cleanup of Tonawanda Laboratory 

facilities, but it was noted at the April 18, 2005, Worker Outreach Meeting that Building 14 was 

contaminated, and later surveys indicated that residual contamination remained after AEC work (BNI 

1993). 

Cleanup of the Ceramics Plant began before the shutdown of Step III production, which occurred on 

June 30, 1949. Some Step II equipment in Building 30 was removed in March and April of 1948 by or 

with the assistance of a contractor (H. K. Ferguson) (Heatherton 1948c,d,e). Dismantling of Step I 

and remaining Step II equipment in Building 30 was under way in May 1949 (Heatherton 1949b). A 

contractor (Kulp Waco) assisted in some of the work (Heatherton 1949c,d).

Shortly after the shutdown, a comprehensive cleanup effort was undertaken to reduce levels of 

radioactivity in Building 30 to enable its release to Linde for unrestricted use (Heatherton 1950). After 

removal of the bulk of the process equipment, the entire building was vacuum-cleaned and flushed 

with water. A systematic radiation survey was conducted to identify areas of contamination. 

Decontamination was accomplished primarily by removing contaminated parts of the building (such as 

portions of the second floor balcony on which process operators had been stationed) and by abrading 

surfaces (mostly by sandblasting, although oxygen acetylene torches were also used).

After each area was decontaminated, it was again cleaned and flushed, and then a final radiation 

survey was made.

Workers who performed surface abrasion activities were provided with U.S. Bureau of Mines (USBM)-

approved rubber hoods that protected the face, neck, and shoulders and had a supplied air respirator 

to permit breathing of uncontaminated air. Other personnel in the area wore demand respirators or, 

for short periods, USBM respirators approved for all dusts.

An AEC internal memorandum from March 29, 1950 reported that the decontamination of Building 30 

had been completed and indicated that a decision on release of the building to Linde was to be made 

by April 1, 1950 (Eisenbud 1950).

Decontamination of Building 38 began with a radiation survey in November 1952 (Harris 1952). An 

AEC memorandum from April 1954 (Harris 1954) reported levels after decontamination. Average 

readings on categories of equipment to be left at the plant ranged from 0.05 to 0.6 mrep/hr beta plus 

gamma with the gamma component either undetectable (which meant less than twice background) or 

less than 5% of the total. The memorandum reported "overall" floor and wall levels to be 

1.01 "mreps/hr/ft2

" beta plus gamma. This was considered excessive. The memorandum cited data 

that indicated this could be reduced to 0.065 mreps/hr/ft2 by covering contaminated floor areas with 

asphalt tile. It recommended release of the building to Linde once this occurred. July 7, 1954, is used 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 27 of 102

as the definitive end of the decontamination period based on the date of the final survey of the facility, 

which was documented in a memorandum from the NYOO to Union Carbide that asserted the 

decontamination requirements of the contract had been fulfilled (Eisenbud 1954).

No information was found on cleanup and decontamination activities in Buildings 31 and 37.

2.7 POST-MED/AEC OPERATIONS

The Formerly Utilized Sites Remedial Action Program (FUSRAP) began in 1976. ORNL surveyed the 

Linde Tonawanda site from October 18 through November 5, 1976, to determine if remediation would 

be required (ORNL 1978). Radiation and radioactive contamination measurements were made inside 

Buildings 14, 30, 31, 37, and 38; on the Tonawanda property outside the buildings; and at nearby 

offsite locations. Linde employees noted that Building 30 renovation occurred in the 1960s and could 

have resulted in elevated employee radiation exposures. Notes from the Worker Outreach Meeting 

on April 18, 2005, mention contamination in association with Building 57, and an additional review of 

the BNI (1993) remedial investigation report shows areas of residual radioactive contamination were 

associated with areas in or near Buildings 57, 58, and 90. The highest indoor radiation levels were 

found in the principal production buildings, Buildings 30 and 38. Linde was designated as a FUSRAP 

site in 1980. Additional radiological surveys and decontamination efforts followed (BNI 1993). These 

led eventually to demolition of Building 14 and all of the Linde Ceramics Plant buildings that were 

involved in MED/AEC work except Building 31. Table 2-1 shows demolition dates. As of 2004, 

Building 31 remained in use and onsite soil remediation was in progress with completion scheduled 

for 2007 (Pilon 2004).
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 28 of 102

3.0

This section develops parameters for reconstruction of doses due to internal exposures from

November 1, 1947, until July 7, 1954. HHS has determined, and NIOSH has concurred that it is not 

feasible to reconstruct internal exposure prior to November 1, 1947 (HHS 2005). 

ESTIMATION OF INTERNAL EXPOSURE, 1947 TO JULY 7, 1954

This section describes information related to internal dose reconstruction including Linde’s uranium 

bioassays, airborne radioactivity concentration measurements, radon breath analyses for 

determination of radium body burdens, and radon air concentrations. To expedite preparation of this 

document, the Linde information is considered in conjunction with information from facilities that did 

similar types of uranium processing to establish preliminary estimates of internal intakes and 

exposures. These estimates are considered best estimates, until data can be further considered. It is 

believed that additional analysis of the data will lower at least some of the intakes and exposures that 

are estimated in this section.

This document assumes that Tonawanda Laboratory primary internal exposures occurred from 

October 1, 1942, to December 31, 1946. At the Ceramics Plant, four periods were assumed: 

October 1, 1942, to July 31, 1946 (operations), August 1, 1946, to September 14, 1947 (standby), 

September 15, 1947, to October 31, 1947 (rehabilitation), and November 1, 1947, to July 7, 1954

(operations and cleanup). From October 1, 1942, into 1946, both African and domestic ore were used 

for research, development, and production. Tonawanda Laboratory AEC work is assumed to have 

stopped radiation work after December 31, 1946, although some workers might have visited the 

Ceramics Plant buildings. The exact levels of contamination that remained at the Ceramics Plant and 

the nature of worker activities in areas of residual activity are unknown for the standby and 

rehabilitation periods. Exposures to uranium progeny and radon would have decreased at the 

Ceramics Plant after August 31, 1946, because ore was no longer processed during this period. 

Continued lower level exposures to uranium progeny and to radon were assumed because some 

radioactive waste was disposed of on the site and because initial cleanup was not completed until the 

end of 1954; however, for the Ceramics Plant the uranium exposures would have dominated during 

the post-1946 period.

The primary sources of internal radiation exposure at Linde due to MED and AEC work were natural 

uranium and its progeny.

3.1 ESTIMATION OF PARTICULATE INTAKES

During World War II and the time of the MED, the permissible level for natural uranium air dust 

concentration was set at 500 μg/m3 for insoluble uranium compounds and 150 μg/m3 for soluble 

uranium compounds. After the war, the University of Rochester in Rochester, New York,

recommended lowering the permissible level to 50 μg/m3 for all uranium compounds on the basis of 

chemical toxicity. This level was also stated as 70 dpm/m3 for natural uranium and was based 

primarily on animal studies. The Medical Division of the AEC NYOO thought that the maximum 

permissible level should be based on human data and was thus unknown. Therefore, the level of 50 

μg/m3 was generally referred to as the preferred level (PL) (AEC 1949a). Contractors of the AEC 

often used the term maximum allowable (air) concentration (MAC) interchangeably with PL and 

reported air-sampling results as multiples of the MAC (Stewart 1952; AEC 1953). When considering 

air concentrations that were reported in multiples of the PL or MAC, the actual assumed value of the 

PL or MAC should be verified.

In 1949, the Medical Division of the NYOO published a report on the health hazards at seven facilities 

that produced and/or processed uranium for the AEC. These facilities included Mallinckrodt Chemical 

Works, Harshaw Chemical Company, LAPC, Electro Metallurgical Company, and Vitro Manufacturing 

Company. The AEC used the information on work tasks with measured air concentrations in 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 29 of 102

breathing zones, general areas, and process areas to determine average air concentrations that were 

weighted by exposure times and summed these time-weighted air concentrations to determine daily 

time-weighted average air concentrations by job category. Up until the time of the 1949 AEC report

(AEC 1949a), surveys by the NYOO indicated that out of 648 exposed workers at these plants, 9% 

were exposed to uranium air concentrations greater than 125 MAC (greater than 6,250 μg/m3

), 9% 

were exposed at 25 to 125 MAC (1,250 to 6,250 μg/m3

), and 82% were exposed to less than 25 MAC 

(less than 1,250 μg/m3

). Linde’s maximum time-weighted exposure during this period was 33 or 

32 MAC (the data in the report’s text and its graph differed). As a result of the NYOO report, 

significant improvements were made in operational conditions such as redesign of ventilation 

systems, enclosing some processes, and using remote controls (AEC 1949a). By the end of 1949, 

exposure levels were significantly reduced at these larger plants even though production levels 

increased (Mason 1958).

Indications are that some of the higher routine (versus episodic) exposures occurred at the uranium 

ore processing facilities. The AEC did not exist until January 1, 1947, and its report (AEC 1949a) did 

not elaborate on internal exposures before its tenure. A general review of air concentration data, 

safety reports, and production and progress reports from the early 1940s through 1946 indicates that 

there were significant reductions in exposures due to improved engineering, process, workplace, and 

administrative controls brought into effect before the AEC report. This indicates that the exposures at 

the ore and waste processing plants in the earliest years of operations would likely have been higher 

than the exposures during AEC tenure. Indeed, the “tolerance” air concentration was 10 times higher 

for insoluble compounds during the early years and 3 times higher for soluble compounds.

After the ore processing, Linde began a standby period. It was assumed that exposures decreased to 

0.1 MAC at the Tonawanda Laboratory after cleanup in 1946 until December 31, 1953. Based on 

reviews of later air concentrations at Linde and reviews of air concentration data from other sites, 

most workers’ exposures would have been much lower during these periods. 

The standby period at Linde Ceramics was assumed to end on September 14, 1947. Rehabilitation of 

the Step III process was assumed to begin on September 15, 1947, and continue through October 31, 

1947. 

3.2 URANIUM

During uranium research and the processing of uranium ores at Tonawanda Laboratory and the 

Ceramics Plant, workers might have been exposed to a variety of uranium chemical forms that 

encompass all lung absorption types: F, M, and S. Although some process steps might have had 

more or less exposure to a given lung absorption type, it is not clear how well separated these areas 

were before standby. In addition, workers could have worked in multiple uranium process areas. This 

analysis assumed that types F, M, and S were available before standby. After operations ceased in 

1946, it is more likely that remaining uranium contamination would have been oxidizing and becoming 

less soluble. Beginning with the rehabilitation period for restart of Step III operations in 1947, it is 

assumed that only type M and S materials would be available. The selection of absorption type 

should depend on the organ of interest. 

3.2.1

Medical urinalyses for specific gravity, proteins, sugars, etc. were a requirement at Linde from the 

beginning of MED work, but there is no evidence that urine was analyzed for radioactivity until 1947. 

MED noted that tentative arrangements had been made for uranium urinalyses of certain groups of 

employees at Linde (Ferry 1944b), but no records are available to indicate the analyses were done. A 

page titled, “Schedule of Examinations, Contract AT-30-1-GEN-165” (Author unknown ca.1947), 

which commenced sometime in 1947 (Rennich 1947), indicates that 60 cm3 of urine from each 

Uranium Urinalysis Data
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 30 of 102

employee was to be sent monthly to the University of Rochester, the location of one of the earliest 

bioassay analysis programs. Individual uranium urinalysis data from November 1947 to 1950 are 

available for some Linde Ceramics workers (LAPC 1947–1950). The period covers both Step III 

production in Building 38 and decontamination in Building 30. Worker radiation exposures during this 

period were likely to be lower than during the production period before standby because African ore 

was no longer being processed, Step I and II processes had ceased, and engineered safety controls 

had been improved.

The reported results of Linde uranium urinalyses during Step III production and later cleanup ranged 

from 0 to 3.10 mg/L, with many of the higher results reported in the November to December 1947 

“preemployment urines.” Because cleanup/preparation activities occurred before collection of these 

samples, it is possible the elevated urinalysis results represented exposures that were received either 

during rehabilitation of the Step III equipment or perhaps from the less soluble component of uranium 

that was taken in during earlier operations. 

In November 1948, two uranium urinalysis results, 3.10 and 0.33 mg/L, were noted as high, and Linde 

was requested to consider the possibility of contamination. A fluoride result from the individual with 

the highest uranium results was also elevated. No information was found to indicate whether these 

were contaminated samples, but the results are consistent with intakes of soluble uranium.

Some samples might have been collected over the weekend: one set of samples indicated that 

collection bottles were taken home on a Friday because there was no work for 2 days, and the bottles 

were returned on Monday morning at 8:00 a.m. A note on the January 1950 uranium urinalysis 

results states, “Please add approx[imately] 1% of conc[entrated] HCl to urine as a preservative 

instead of toluene.” 

It appears that the University of Rochester and the AEC NYOO were performing the uranium 

urinalyses for Linde (Wolf 1948). The uranium fusion photofluorimetry urinalyses that were performed 

by the University of Rochester and the AEC NYOO were similar to those at other AEC facilities. The 

default detection threshold for uranium urinalysis was assumed to be 10 μg/L based on a reported 

sensitivity of 5 to 10 μg/L for uranium fluorimetry urinalysis in the early years (Wilson 1958). However, 

the January 1948 uranium urinalysis results AEC transmitted to Linde included a note that indicated

the minimum detectable amount might have been much higher (Wolf 1948):

The uranium content of those samples listed as being 0.1 mg U/l or less is below the 

limit of accurate quantitative determination by the photofluorometric method of 

analysis. 

February, May, June, August, and November 1948 uranium urinalysis results included a similar note, 

but 0.01 mg U/L was reported as the limit of accurate quantitative determination or reliable 

determination for photofluorometric analysis. The January 1948 determination level of 0.1 mg/L is 

assumed to be a typographical error because this is the same as the determination level reported for 

(nonradioactive) fluoride urinalysis and because there seems to be no change in the format of the

reported numbers. 

The AEC requested “check samples” on all uranium urinalyses that exceeded 0.1 mg/L, and noted in 

February 1948 that it would be appropriate to resample in about 3 months (Wolf 1948). Linde’s reply 

requested clarification because it had previously understood that resampling was to occur monthly

(Heatherton 1948f).

Analysis of Coworker Bioassay Data for Internal Dose Assignment (ORAU 2005d) describes the 

general process used for analyzing bioassay data for assigning doses to individuals based on

coworker results. Bioassay results described above were analyzed in accordance with this procedure 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 31 of 102

(Attachment D). The results of this analysis are presented in Tables 3-1 and 3-2. Individual uranium 

urinalysis results should be used to determine internal exposure to the individual when they are 

available. Where individual results are not available, the coworker data included in Attachment D and 

summarized in Tables 3-1 and 3-2 are to be used to estimate internal exposures that are favorable to 

claimants.

Table 3-1. Chronic intake rate for Type M uranium (pCi/d).

Start date End date

50th-percentile

value GSD

11/01/1947 07/07/1954 74 4.0

Table 3-2. Chronic intake rate for Type S uranium (pCi/d).

Start date End date

50th-percentile

value GSD

11/01/1947 07/07/1954 1884 4.3

3.3 RADIUM

HHS has determined, and NIOSH has concurred that it is not feasible to reconstruct internal exposure 

prior to November 1, 1947 (HHS 2005). Information on radon exposure prior to November 1, 1947 is 

provided only as a basis for extrapolation afterwards and is not intended to be used during the period 

in which reconstruction of internal dose has been determined to be infeasible. 

All radium compounds are lung absorption type M. Radon breath analyses have been used to 

provide information on the amount of radium in the body and are available for some Linde workers. 

Radon Breath Data

The level of radon in exhaled breath can provide information on the amount of radium in the body. 

Radon breath data were found for 45 samples from Linde workers between June 12, 1944, and 

January 17, 1945 (LAPC 1944–1945). The records indicate that Dr. R. D. Evans in Cambridge, 

Massachusetts, analyzed some of the samples. Room background concentrations of radon were 

sometimes subtracted from the results and sometimes not. In July and August 1944, room 

background concentrations were reported respectively as 0.6 and 0.3 pCi/L (Tybout 1944b). The 

measured radon levels were given for some but not all of the samples. Reported radon breath results 

ranged up to 2.2 pCi/L. The method of deriving the 226Ra level was not stated, but in many cases, the 

reported burden of radium in micrograms was numerically equal to 10% of the breath radon 

concentration in picocuries. Radon breath results are the starting point for dose reconstruction, so it 

might be necessary to back-calculate breath results from either 226Ra burdens or tolerance levels. 

The original records might have sufficient information for this determination. This information has not 

yet been tabulated for inclusion in this site profile.

3.4 URANIUM PROGENY

Ceramics Plant 1943 to 1946 Production, and Tonawanda Laboratories

HHS has determined, and NIOSH has concurred that it is not feasible to reconstruct internal exposure 

prior to November 1, 1947 (HHS 2005). 

Ceramics Plant 1947 to 1949 Step III Production and Subsequent Initial Cleanup

During this period, refined uranium materials were handled. None of the uranium progeny would have 

been present in significant quantities in the refined uranium materials but, to account for uranium 

progeny potentially present from past activities and resuspended during decontamination and 

decommissioning (D&D) activities, data from the postoperations period was reviewed to determine 

bounding activity ratios (Attachment E). Table 3-3 presents bounding indoor uranium progeny ratios 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 32 of 102

for use for dose reconstruction for the period from November 1, 1947, through July 7, 1954. The 

values in this table were the highest observed values from the indoor and storm sewer sampling 

locations.

Table 3-3. Progeny to uranium ratios.

Progeny/U (total) Ratio to uranium

Th-230/U 0.26

Ra-226/U 0.21

Po-210/Ua 0.21

Ac-227/U 0.29

Pa-231/U 0.01

a. Po-210 activity not reported, assumed to be the 

same as parent (Ra-226)

3.5 RADON

HHS has determined, and NIOSH has concurred that it is not feasible to reconstruct internal exposure 

prior to November 1, 1947 (HHS 2005). Information on radon exposure prior to November 1, 1947 is 

provided only as a basis for extrapolation afterwards and is not intended to be used during the period 

in which reconstruction of internal dose has been determined to be infeasible. 

More than 200 measurements of radon concentrations were made during 1942 to 1946 pilot plant and 

production processing of African ore at Linde (LAPC 1944–1946). The early measurements were 

made using liter glass bulbs that were evacuated immediately before sampling to below 50 μm 

(Skinner 1944). It is likely that the later measurements during operations were made in the same 

way. Early samples were analyzed at the Massachusetts Institute of Technology; later analyses 

appear to have been performed at the University of Rochester. The tolerance value for radon, 

sometimes abbreviated as TV in the old records, was 100 pCi/L.

To determine exposure due to a specified radon concentration (assumed to be 222Rn), the 

concentration C in picocuries per liter was converted to potential alpha energy concentration (PAEC) 

in units of working level (WL) using:

PAEC = C × F ÷ (100 pCi/L per WL) (3-1)

where the equilibrium factor F was taken as 0.4, a value that is recommended by International 

Commission on Radiological Protection (ICRP) Publication 32 (ICRP 1981) and the United Nations 

Scientific Committee on the Effects of Atomic Radiation (UNSCEAR 1993). Potential alpha energy 

exposure (PAEE) in units of working level-months (WLMs) was obtained from the PAEC and the 

number of months of exposure M using:

PAEE = PAEC × M (3-2)

where a month is assumed to be 170 work hours. 

3.5.1

During Ceramics Plant preproduction and initial production (which involved only domestic ore 

processing), the only source of radon was African ore processing at Tonawanda Laboratory. The 

indoor and outdoor radon concentrations to which Ceramics Plant workers were exposed were 

assumed to equal the outdoor concentration from Tonawanda Laboratory work. No direct 

measurement of this was available. An estimate was made based on the lowest measured indoor 

concentrations at the Ceramics Plant during African ore processing. These were viewed as indicating 

Ceramics Plant Radon Exposures
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 33 of 102

the upper limit to the outdoor concentration because outdoor air was drawn indoors for ventilation. 

Approximately 20% of the measurements in the Ceramics Plant ore processing building yielded 

results of 10 pCi/L or less, with most of these results at or near 10 pCi/L. Therefore, 10 pCi/L was 

taken as the estimated outdoor concentration.

Measured radon concentrations during processing of African ore at the Ceramics Plant are 

summarized in Table 3-4. Based on geometric means (GMs), locations were grouped into three 

categories – high, medium, and low – and the GM and geometric standard deviation (GSD) of all 

results in each category were determined. Some results were reported as less-than values and these 

are denoted as less than the limit of detection (<LOD). 

Table 3-4. Measured radon concentrations during Ceramics Plant 

African ore processing.

Location category/location

# of 

samples

GM

(pCi/L)a GSD

GM

(pCi/L)a GSD

High 99.3 3.43

Scales 5 153 2.50

Ore box car 17 125 3.92

Ore sampling room 24 100 3.11

Ore dumping grill 18 88.6 2.91

Ore storeroom 26 88.5 4.49

Ore bin/conveyor 7 83.3 9.37

Medium 42.5 2.72

Ore bag tumbler 17 60.2 2.53

Ore bag wash room 8 45.1 2.78

ball mill & classifier 24 32.5 2.74

Low 22.4 3.17

Step 1 digestion & filtration 55 23.7 3.17

Loading dock 1 18.0 (b)

Receiving desk 5 12.9 2.43

a. In computing the GM, a <LOD value was assumed to equal the LOD.

b. GSD could not be determined.

After the end of African ore processing, concentration in the main ore processing building, Building 30, 

was assumed to remain at the 10-pCi/L level that was measured during the second period of domestic 

ore processing until the end of cleanup of Building 30. Concentrations in other Ceramics Plant 

buildings were also assumed to be 10 pCi/L until the end of cleanup in those buildings. Because the 

locations of many workers are likely to be unknown, it was assumed that all workers were exposed to 

10 pCi/L of radon from November 1, 1947, through July 7, 1954. Table 3-5 lists concentrations and 

exposure rates.

Table 3-5. Ceramics Plant worker radon exposures rates, 1947 to 1954.

Period/work location

Time-weighted

concentration (pCi/L)

Exposure rate

(WLM/yr)

11/01/1947–07/07/1954

All workers 10.0 0.480

3.5.2

Few measurements were available for the Tonawanda Laboratory, so estimates of radon 

concentrations were based on Ceramics Plant data. African ore work at Tonawanda Laboratory 

appears to have occurred just before each period of African ore processing at the Ceramics Plant. 

The precise dates of African ore research at Tonawanda Laboratory are not known. Based on the 

available reports, it appears to have occurred only about half of the time that the Laboratory was 

engaged in MED work, with most of the African ore work concentrated in the periods immediately 

Tonawanda Laboratory Radon Exposures
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 34 of 102

before the beginnings of the two African ore processing periods at the Ceramics Plant. For dose 

reconstruction, it was assumed that African ore processing at Tonawanda Laboratory occurred during 

the entire MED period, but that the peak concentrations of radon were equal to the medium values at 

the Ceramics Plant. It was assumed that after MED research and initial cleanup at the Laboratory 

ended, the radon concentration dropped to 10 pCi/L (the level in the Ceramics Plant after it switched 

from African ore processing to domestic ore processing) and remained there until the end of the 

cleanup at the Laboratory. After the end of cleanup, radon exposure in the Laboratory was based on 

the highest GM PAEC that was determined for a Tonawanda site building from measurements in 

1981. That value was 1.68 × 10-02 WL for Building 31 based on analysis of data in BNI (1982, 

Table B-3). Table 3-6 summarizes the assumed radon concentrations and resultant exposures.

Table 3-6. Tonawanda Laboratory radon exposure rates, 1947 to 1953

Period Time-weighted 

concentration (pCi/L)

Exposure rate

Start End (WLM/yr)

Postcleanup

11/01/1947 12/31/1953 — 0.202

Radon and Radon Progeny

The radon exposure rates that are provided here are annual PAEE rates in WLM/yr. Each value is 

assumed to be the median of a lognormal distribution with a GSD of 3.43. This GSD is based on the 

location category having the highest GSD (3.43 per Table 3-4). 

3.6 INHALATION INTAKE ESTIMATES OF RADIOACTIVE PARTICULATES AT 

TONAWANDA LABORATORY

The airborne concentrations in Section 3.1 can be used to determine uranium intakes for Tonawanda 

Laboratory employees from November 1, 1947, through December 31, 1953. This analysis assumed

that a worker was chronically exposed to alpha airborne concentrations as stated in Section 3.1 and 

summarized in Table 3-7. 

Table 3-7. Assumed airborne concentrations used to estimate intakes at the Tonawanda Laboratory. 

Start End Activity description # MAC α dpm/m3 Source

11/01/1947 12/31/1953 Postcleanup 0.1 7 Uranium and progeny

This analysis assumed Linde workers were exposed to the given air concentrations for 8 hr/d, 5 d/wk, 

and 50 wk/yr. Time-weighted exposure studies of even the larger plants like Mallinckrodt showed that 

the majority of workers were exposed at lower levels than were estimated for the early 1942 to 1946 

operational period. No credit is taken for breaks or working at tasks where radioactive material 

intakes would be much lower or nonexistent, or for the use of facemasks; on the other hand no 

consideration was given to the longer workweeks and work hours in the early years. In addition, 

because workers were exposed to multiple absorption types, but the single absorption type that 

produces the larger dose is used to estimate organ dose, it is believed that this estimate adequately 

accounts for internal exposure. 

Because the early air concentration measurements would have included detection of radiations from 

progeny, the air concentrations that are listed in Table 3-7 as including progeny are apportioned 

based on the assumption of full equilibrium among uranium and its progeny, using the alpha ratios of 

uranium and progeny of internal dosimetric importance. 

For example, the annual uranium inhalation intake due to chronic exposure at 0.1 MAC was estimated 

by multiplying the air concentration of 7 dpm/m3 by the alpha fraction of uranium (0.489), the ICRP 

Publication 66 (ICRP 1994) recommended breathing rate of 1.2 m3

/hr, and the assumed 2,000 work
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 35 of 102

hours per calendar year. This results in an annual chronic inhalation intake of 8.215 × 103 dpm, which 

is equal to a daily intake rate of 22.5 dpm/d.

Inhalations of the uranium progeny were estimated by substituting the alpha fraction of the progeny 

for the alpha fraction of the uranium. 

3.7 INGESTION INTAKE ESTIMATES AT TONAWANDA LABORATORY

In the case where inhalation intakes are calculated from air concentrations, ingestion intakes are also 

to be considered. NIOSH (2004) indicates that the ingestion rate, in terms of dpm for an 8-hour 

workday, can be estimated by multiplying the air concentration in dpm per cubic meter by a factor of 

0.2, so the uranium ingestion rate based on an air concentration of 7 alpha dpm/m3 would be 

0.563 dpm/wd. To adjust this to ingestion intake per calendar day, 0.685 dpm/wd was multiplied by 

250 wd/yr and divided by 365 d/yr, which equals 0.469 dpm/d. In accordance with NIOSH (2004), the 

f1-value used for inhalation dose calculations is to be used for ingestion dose calculations.

3.8 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

HHS has determined, and NIOSH has concurred that it is not feasible to reconstruct internal exposure 

prior to November 1, 1947 (HHS 2005). 

Intake estimates and radon exposures that were estimated for Linde workers are shown in the Tables 

3-8, 3-9, 3-10, and 3-11. Uranium intakes are based on the coworker data contained in Tables 3-1 

and 3-2. Uranium progeny intakes were determined by applying the isotopic ratios in Table 3-3. For 

the ceramics plant, either the values in Table 3-8 or 3-9 should be applied (not both), whichever is 

most favorable to claimants along with the radon intake in Table 3-10. Tonawanda laboratory 

employees should be assigned the intakes in Table 3-11. For workers whose work location is 

considered indeterminate, intakes for the Ceramics Plant should be assumed. 

Uranium bioassays are available for some workers during the period from November 1947 through 

January 1950. If uranium bioassays are used to reconstruct an individual's dose (as opposed to the 

values tabulated below), additional intakes from uranium progeny must be added using the activity 

ratios in Table 3-3.

Table 3-8. Internal exposure summary for the Ceramics Plant, November 1, 1947, to July 7, 1954

(based on Type M uranium).

Radionuclide Start End

Intake

route

Absorption

type

Intake or 

exposure Units

U-234 11/01/1947 07/07/1954 Inhalation M 1.64E+02a dpm/d

Th-230 11/01/1947 07/07/1954 Inhalation M 4.26E+01a dpm/d

Ra-226 11/01/1947 07/07/1954 Inhalation M 3.44E+01a dpm/d

Po-210 11/01/1947 07/07/1954 Inhalation M 3.44E+01a dpm/d

Pa-231 11/01/1947 07/07/1954 Inhalation M 1.64E+00a dpm/d

Ac-227 11/01/1947 07/07/1954 Inhalation M 4.76E+01a dpm/d

a. Intake assuming type M uranium, value listed is GM of a lognormal distribution with GSD of 4.0.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 36 of 102

Table 3-9. Internal exposure summary for the Ceramics Plant, November 1, 1947, to July 7, 1954

(based on Type S uranium) .

Radionuclide Start End

Intake

route

Absorption

type

Intake or 

exposure Units

U-234 11/01/1947 07/07/1954 Inhalation S 4.18E+03a dpm/d

Th-230 11/01/1947 07/07/1954 Inhalation S 1.09E+03a dpm/d

Ra-226 11/01/1947 07/07/1954 Inhalation M 8.78E+02a dpm/d

Po-210 11/01/1947 07/07/1954 Inhalation M 8.78E+02a dpm/d

Pa-231 11/01/1947 07/07/1954 Inhalation S 4.18E+01a dpm/d

Ac-227 11/01/1947 07/07/1954 Inhalation S 1.21E+03a dpm/d

a. Intake assuming type S uranium, value listed is GM of a lognormal distribution with GSD of 4.3.

Table 3-10. Internal exposure summary for the Ceramics Plant, November 1, 1947, to July 7, 1954.

Radionuclide Start End

Intake

route

Absorption

type

Intake or 

exposure Units

Rn-222 11/01/1947 07/07/1954 Inhalation (a) 0.48E+00 WLM/yr

a. Value shown is for full year, partial years should be prorated.

Table 3-11. Internal exposure summary for the Tonawanda Laboratory, November 1, 1947, to

December 31, 1953.

Radionuclide Start End

Intake

route

Absorption

type

Intake or 

exposureb Units

U-234 11/01/1947 12/31/1953 Inhalation M, S 2.25E+01 dpm/d

Ingestion (a) 4.69E-01 dpm/d

Th-230 11/01/1947 12/31/1953 Inhalation M, S 1.10E+01 dpm/d

Ingestion (a) 2.29E-01 dpm/d

Ra-226 11/01/1947 07/07/1954 Inhalation M 1.10E+01 dpm/d

Ingestion (a) 2.29E-01 dpm/d

Po-210 11/01/1947 07/07/1954 Inhalation F, M 1.10E+01 dpm/d

Ingestion (a) 2.29E-01 dpm/d

Pa-231 11/01/1947 07/07/1954 Inhalation M, S 5.13E-01 dpm/d

Ingestion (a) 1.07E-02 dpm/d

Ac-227 11/01/1947 07/07/1954 Inhalation F, M, S 5.13E-01 dpm/d

Ingestion (a) 1.07E-02 dpm/d

Rn-222 11/01/1947 12/31/1953 Inhalation - 2.021E-01 WLM/yr

a. Choose same f1-value as used for inhalation per NIOSH (2004).

b. The dose distribution for particulate intakes is assumed to be constant. The dose distribution for radon is assumed to be 

lognormal with a GSD of 3.43.
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4.0

Because of the SEC determination (HHS 2005) that it is infeasible to adequately reconstruct internal 

dose during the period October 1, 1942 through October 31, 1947, dose estimates for this period are 

considered partial dose estimates.

ESTIMATION OF EXTERNAL EXPOSURE, 1942 TO JULY 7, 1954

This section develops parameters for reconstruction of doses due to external exposures due to work 

activities from October 1, 1942, the assumed start date of MED work at Linde, until December 31, 

1953. In some parts of this section, measurements and parameters for the period after 1953 are 

analyzed because they provide information used to estimate external exposures before 1954. 

Occupational medical exposures are treated in Section 5.0. Dose reconstruction for the period after 

1953 is discussed in Section 6.0. Film badges were worn by some workers during some periods. For 

unmonitored workers and periods, doses are estimated from source term information, workplace 

measurements, and from available dosimetry results for workers.

Throughout this document, it is assumed that 1 mR = 1 mrad(air) = 1 mrad(tissue) = 1 mrem = 

1 mrep. 

Beta Radiation

In developing the beta dose rate estimates in this section, no attenuation due to shoes, apparel, or 

protective items (e.g., gloves, face shields) was considered. Attenuation due to apparel is variable 

and sometimes not very significant (see Table A-5). The types of protective items issued to workers 

varied with time, and the extent to which workers used them is uncertain.

Workers who frequently handled significant quantities of beta-emitting materials were assumed to 

have had higher beta doses to the hands and forearms than to the remainder of the body. The 

relationships assumed between contact dose rate and dose rates to other parts of the body were 

based in part on data from the Ceramics Plant (see Table A-6). 

For the purpose of calculation of organ dose, all exposure geometries are assumed to be anteriorposterior (AP). 

Burlap Bag Exposures

Based on the assessment contained in Attachment E of this TBD, NIOSH ' s review of historical 

records indicates that the period during which the burlap bags were staged and burned is within the 

current SEC period. NIOSH has found no other credible evidence, other than a former employee 

affidavit which indicated that two pallets of filled burlap bags were in Building 30 at Linde Ceramics, to 

support a potential burlap bag exposure issue outside of the SEC period. If credible information is 

presented in an individual CATI, or documentation related to an individual claim, to establish the 

worker’s occupational proximity to the burlap bags, dose reconstructors should use the external dose 

estimates provided in Attachment E to assess this external exposure scenario to support assessment 

of this external dose for the purpose of individual dose reconstruction.

4.1 CERAMICS PLANT BETA AND GAMMA EXPOSURE

4.1.1

A systematic characterization of radiation levels in Building 30 was made in 1949 in conjunction with 

its decontamination (Heatherton 1950). A survey scheme called “restricted randomization” was used. 

The floor area was divided into 15- by 15-ft squares each of which in turn was divided into 100 small 

squares. In each large square, four of the small squares were selected at random and surveyed. 

Preproduction, 1942 to 1943
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Total beta plus gamma radiation in millirep per hour was measured at contact at the square’s center 

and corners and at 3 ft above the floor. A similar method was used to measure wall contamination. 

Before the initial survey, the building was vacuumed and flushed with water. After the initial survey, 

heavily contaminated portions of the building were removed or cleaned, mostly by sandblasting. Then 

the building was again vacuumed and flushed, and a final survey was taken. Heatherton (1950) 

reported the number of measurements and the lowest, mean, and highest values for each of several 

major plant areas. The middle columns of Table 4-1 display the results of a statistical analysis of the 

data. 

Another comprehensive survey of Building 30 was made in 1976 (ORNL 1978). Beta plus gamma 

levels in millirad per hour were measured at 1 cm from the surface, and gamma levels microroentgen 

per hour were measured at 1 m. Table 4-2 displays the results of a statistical analysis of the data. 

The measured levels were similar to the postdecontamination results of 1950, which indicated little 

change in conditions. Because gamma measurements were reported, the beta and gamma 

percentages of the radiation could be estimated. These were used to estimate the beta and gamma 

components in Table 4-1 for the pre- and postdecontamination results.

Table 4-1. Floor and wall radiation in Building 30 before and after the 1949 to 1950 decontamination.

Time Location

Measured beta plus gammaa Estimated

Median (mrep/hr) GSD Beta (mrem/hr) Gamma (mR/hr)

Before vacuum cleaning 

and flushing

Contact (b) (b) 2.03c 0.131c

3 ft (b) (b) 0.676c 0.131c

Predecontamination (just 

before decontamination 

and after vacuum 

cleaning and flushing)

Contact 0.719 1.67 0.675d 0.0438d

3 ft 0.240 1.67 0.225d 0.0438e

After decontamination Contact 0.311 2.04 0.292d 0.0189d

3 ft 0.0940 2.04 0.0883d 0.0189e

a. Based on data in Heatherton (1950, Tables III and IV). The data were assumed to be characterized by a lognormal 

distribution. For the contact measurements, the mean value and a percentile value were available (the percent of 

results less than 1 mrep/hr was stated). These were used with the computer program LOGNORM4 (Strom and 

Stansbury 2000) to estimate the median and GSD. For measurements at 3 ft, the mean was available but not the 

percentile value. The GSD determined from the contact data was assumed to apply and used to estimate the median.

b. Not available.

c. Assumed to be three times higher than the predecontamination values. The factor three is based on April 1949 data; 

see discussion in text.

d. Estimates based on beta and gamma percentages in Table 4-2.

e. Assumed to equal contact gamma exposure rate.

Table 4-2. Building 30 floor radiation levels in 1976.

Parameter

GMa

GSDa Composition of 

Value Units radiation at 1 cm

Beta plus gamma at 1 cm 0.253 mrad/hr 3.47 (b)

Gamma at 1 mc 0.0154 mR/hr 1.69 (d)

Estimated gamma at 1 cme 0.0154 mR/hr (d) 6.09%

Estimated beta at 1 cmf 0.238 mrad/hr (d) 93.9%

a. GM and GSD were calculated from data in Table 3 of ORNL (1978).

b. Estimate made in last two lines of the table.

c. Background was not subtracted from the external gamma values. Outdoor external 

background in the Tonawanda area was reported to be 8 to 15 μR/hr (ORNL 1978, 

p. 15).

d. Not applicable.

e. Estimated as equal to gamma at 1 m.

f. Estimated as 0.253 (beta plus gamma) less 0.0154 (gamma).

The measurements just before decontamination in 1949 and 1950 were made after the building had 

undergone vacuum cleaning and flushing. Brief, semiquantitative reports were available for two 1-day 
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surveys in April 1949 before the vacuum cleaning and flushing (Blatz 1949; Wolf 1949a). Typical 

levels of measured gamma radiation at 3 ft on April 19 and at contact on April 22 were similar and 

averaged about 0.18 mR/hr, about 4 times higher than the estimated median contact gamma level 

before decontamination. Typical levels of measured beta at contact on April 22 averaged about 

0.87 mrep/hr, about 1.3 times higher than the estimated median beta level before decontamination. 

The April results were the basis of the estimate in Table 4-1 that beta and gamma radiation levels 

before vacuum cleaning and flushing were 3 times higher than the values that were measured 

afterward.

In the 1976 survey, Building 30 was the most contaminated building on the site (ORNL 1978). The 

levels for Building 30 were assumed to apply to all buildings on the site. 

For the preproduction period, it was assumed that there was no significant external beta or gamma 

exposure indoors for Ceramics Plant workers because the buildings were under construction or new. 

Some exposure while outdoors was possible because the transport of ore to Tonawanda Laboratory 

(Building 14) and the work there could have produced some site contamination. Resultant outdoor 

radiation levels are very likely to have been less than floor and wall radiation levels inside Building 14 

because the indoor and outdoor sources would have been comparable but the outdoor sources could 

spread over a larger area and were subject to dispersal by the weather. The radiation levels before 

vacuuming and flushing that are estimated in Table 4-1 are taken as an estimate of the levels in 

Building 14 and also as an upper limit estimate of site outdoor radiation levels. Ceramics Plant 

workers were assumed to have been exposed to these levels for 0.5 hr/d. Table 4-3 shows estimated 

annual exposure rates. 

Table 4-3. Preproduction beta and gamma radiation levels.

Beta (rem/yr)

a Gamma (R/yr)a GSDb

1.01E-01 1.97E-02 3

a. Based on Table 4-1 beta and gamma levels at 3 ft before vacuum 

cleaning and flushing. Assumed exposure period: 0.5 hr/wd, 6 

wd/wk, 50 wk/yr. The beta and gamma rates are for the whole 

body.

b. Estimated on the basis of the GSD values in Tables 4-1 and 4-2.

The beta dose rate was assumed to apply to the whole body with no added dose to the hands and 

forearms because it is unlikely that there was any significant handling of radioactive materials by 

Ceramics Plant workers in this period.

4.1.2

4.1.2.1 Beta 

Production, 1943 to 1946

No beta film badge data were found for 1943 to 1946. There were beta film badge data for 1947 to 

1949, but none of the data pertained to Steps I or II, as these processes were no longer in operation. 

For 1943 to 1946, one report of workplace beta exposure rates was found. Solid samples from the 

Ceramics Plant were sent on January 23, 1944, to laboratories of the Medical Section of the MED for 

measurement of beta radiation. Results were reported as “Beta Radiation Roentgens/8 Hr Day" 

(Ferry 1944c) and are reproduced here in the first five columns of Table 4-4. The roentgen is 

technically defined only for photon radiation; beta dose should be expressed in absorbed dose units 

such as rad or rem. However, because there is an approximate numerical equivalence between 

these units, it is assumed here that the roentgen unit used is equivalent to shallow dose at 0.07 mm in 

rem.

From the fact that no distance values were reported with the measurements and from later discussion 

of the measurements in MED correspondence (Thomas 1944), it is apparent that contact exposure 
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rates were reported. This is supported by the similarity of the results to contact dose rates for various 

natural uranium materials (see Table A-1). 

The fourth and fifth columns of Table 4-4 contain the allowable times of exposure based on a 

3.0-R/wk tolerance level and the actual hours of exposure per week as stated in the MED report. For 

the L-30 tailings sample, the actual exposure time was stated to be 50 hr/wk, much greater than the 

allowable exposure time of 18 hours. A MED memorandum (Hadlock 1944) questioned this 

discrepancy. It brought the following MED response (Thomas 1944):

No action has been taken on the Captain Ferry's recommendation that the exposure 

time to L-30 tailings be limited to 18 hours. The primary reason for this is the fact that 

the estimated fifty hours is based on the time which a man is actually within range of 

the radiation whereas the Medical Section's allowable exposure time is based on 

contact of the tailings directly with the hands....

Table 4-4. Beta radiation exposure rates measured at Linde Ceramics, January 23, 1944.

From MED report of data during processing of L-30 African orea Estimated 

worker 

dose rate 

(rem/yr)d Sample # Sample location

Beta radiation 

(R/8-hr day)b

Allowable time of 

exposure per 

worker (hr/wk)c

Actual time of 

exposure per 

worker (hr/wk)

13 L-30 Ore, Step I 0.1 150 50 31

9 L-30 Tailings, Step I 1.4 18 50 263

12 Barium Cake, Step I 0.8 30 0.33 2

14 A. L. Cake, Step I 1.3 18 9 73

15 PbS Cake, Step I 0.2 120 2 3

16 Soda Salt, Step I 0.5 48 9 28

10 Iron Cake, Step II 0.8 30 3 15

11 N.G. Cake, Step II 4.2 6 6 158

a. Results reported by the MED for measurements made on solid samples sent from Linde Ceramics January 23, 1944 

(Ferry 1944c). 

b. Units are those stated in the MED report.

c. The MED allowable time of exposure was based on the then-effective tolerance levels of 0.5 R/d and 3.0 R per 6-day 

week for beta radiation.

d. Except for Sample #9 (L-30 tailings), the dose rate is based on an assumed 50-week year and the dose rate and actual 

hours of exposure per week stated by the MED. For the worker handling L-30 tailings, the dose rate is estimated based 

on information in the MED letter (Thomas 1944) cited in the text. The dose rate when the worker is "within range of the 

radiation" is estimated as half of the contact dose rate. If the source is sufficiently large and there is no shielding, a 

distance of 1.6 m from the tailings would be required to reduce the dose rate by a factor of 2 (see Table A-4). The dose 

estimate calculation is as follows:

Location (rem/8hr) hr/wk hr/yr rem/yr

Hands in tailings 1.4 10 500 88

Within range of radiation 0.7 40 2,000 175

Total 263

The Linde operating and safety sections have estimated that the average time a worker 

spends with his hands in the tailings is six to eight hours one week. The maximum 

time is never over ten hours one week and even then heavy rubber gloves are worn. 

The last column of Table 4-4 shows worker doses derived from the data. These range from 2 to 

263 rem/yr.

It is striking that the 1943 to 1946 dose rates in Table 4-4 are much higher than the 1947 to 1949 

rates Section 4.1.4.1 discusses. For 1943 to 1946 Step I process operators, the beta dose rate was 

estimated as 263 rem/yr. For 1947 to 1949, the job with the highest beta dose rate was that of a 

Step III process operator, and film badge data indicate that the beta dose rate was only about 
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2 rem/yr (approximate value of annual DD + DM; see Table 4-16 later in this section). Despite the high 

estimates, the 1943 to 1946 data does not appear unreasonable. The dose rates in Table 4-4 are 

typical of contact dose rates for uranium materials (see Table A-1). The worker exposure times were 

debated and reviewed within the MED. In the MED’s interpretation of the data, with allowances for 

the protective measures (e.g., gloves) the rates were within the limit in effect at the time, 3.0 R per 

6-day week (Ferry 1944c) or 150 R/yr. Therefore, the estimates in Table 4-4 are considered a valid 

basis for dose reconstruction.

4.1.2.1.1 Estimation of Worker Doses

Beta doses for specific jobs are estimated below and summarized in Table 4-24 at the end of this 

section.

1943 to 1946 Loaders, Step I Process Operators, Step II Process Operators, Ball Mill Operators, 

and Weighmasters

The calculated dose in Table 4-4 for sample 13 was assumed to apply to a loader. The doses for 

samples 9 and 11 were assumed to be the doses to the most highly exposed Step I and Step II 

process operators, respectively. The dose for the most highly exposed process operator was 

assigned to all process operators in a step. A ball mill operator was assigned the same dose as a 

Step I process operator because the ball mill operators performed some of the tasks of process 

operators when not operating the ball mill (see job description for CL-3, Chemical Operator C on p. 13 

of LAPC 1945b). Reductions for shielding by apparel and containers were not taken into account in 

estimating doses because of insufficient information. The dose rates in Table 4-4 are based on 

contact measurements, so they represent dose rates to the hands and forearms. For all but the 

Loader, dose rates to the remainder of the body were taken to be one-third of the contact dose rates. 

This reduction was based on data in Table A-6 (see Hands and Forearms Dose in Section 4.1.4.1). 

For the loader, dose rates for the remainder of the body were assumed to equal the contact dose 

rates because Loaders frequently hand carried ore bags, pushed carts of ore, and worked in the close 

vicinity of large ore piles.

The weighmaster was responsible for verifying weights of all materials (incoming, product, and 

byproduct; see p. 12 of LAPC 1945b) and so had frequent close contact with these materials. 

Furthermore, a radiological survey data sheet from March 1944 indicates that at that time the 

weighmaster was stationed for 8 hr/d at a location 4 ft from a pile of ore bags. The data in Table 4-4

and Table A-1 indicate that processed materials and byproducts had about 15 times the beta dose 

rate of ore. Therefore, the weighmaster was assigned the same beta exposure rate as a Step I 

process operator.

The dose rates in Table 4-4 were based on measurements on a particular batch of African L-30 ore. 

For a given type of ore, beta dose rates to ore handlers in Step I would be approximately proportional 

to ore grade (the weight percent of U3O8 in the ore). Ore type (African or domestic) would also matter. 

For the fully prerefined ore (which the domestic ore used at Linde might have approximated) the 

electron energy release per uranium decay is only 39% of that for African ore (compare Table A-2 and 

Table A-3). Beta dose rates in Steps II and III would be independent of ore grade. Dependence on 

ore type should have been weak because most radioactive impurities would have been removed in 

Step I.

To use the data in Table 4-4 to estimate time-averaged beta dose rates applicable to the entire 1943

to 1946 production period at Linde, the ratio of the average dose rates to those in Table 4-4 was 

estimated. Table 4-5 documents the determination of the ratio. The grade range of the L-30 ore that 

was used at Linde was 8% to 12% (Aerospace Corporation 1981, Table B-1). To obtain the highest 

ratio, it was assumed that the measurements in Table 4-4 were made on the lowest L-30 ore grade, 

8%. It was also assumed that beta dose was proportional to electron energy that is released per 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 42 of 102

decay and that the worker doses were proportional to the mass of ore processed. With these 

assumptions, it was estimated that average doses would have been 0.84 of the doses that would be 

predicted by using the data in Table 4-4. Therefore, the estimated dose rates in Table 4-4 were 

multiplied by 0.84 to obtain time-averaged dose rates for 1943 to 1946 production. The results are in 

Table 4-6, which summarizes the results of all beta dose rate estimates in this section and divides the 

job categories into three groups (high, medium, and low).

1943 to 1946 Workers with Jobs Analogous to 1947 to 1949 Jobs

Workers in 1943 to 1946 who held jobs analogous to ones in 1947 to 1949 were assigned 3 times the 

1947 to 1949 beta dose rates. The factor of 3 increase accounts for potential exposure to radiation 

from waste products from unrefined uranium ore and for the possibility that procedures in 1943 to 

1946 did not involve as much radiological protection. (Tables A-2 and A-3 indicate unrefined uranium 

materials release approximately 2.6 times as much electron energy per uranium decay as refined 

uranium materials. In addition, the plant had a health physicist in 1947 to 1949 but not in 1943 to 

1946.) Workers with jobs analogous to those assigned to the medium beta exposure category for 

1947 to 1949 were assigned dose rates of 17.6 rem/yr to the hands and forearms and 5.85 rem/yr to 

the remainder of the body. 

Table 4-5. Dependence of beta dose rate on ore type.

Approximate 

processing 

perioda

Ore 

IDa

Ore 

typea

Electron 

energy 

released 

per decay 

(MeV/nt)b

Average 

gradea

Average 

beta dose 

rate 

relative to 

8% L-30c

Mass 

processed 

(MT)a

Mass-weighted 

average beta 

dose rate

relative to 8% 

L-30d

06/1943 to 11/1943 Not 

stated

Domestic 

& scrap

0.447 15.0% 0.72 1,000 0.028

12/1943 to 10/1944 L-30 African 1.157 10.8% 1.35 8,504 0.447

10/1944 to 11/1944 L-50 African 1.157 6.7% 0.84 1,486 0.048

12/1944 to 02/1946 L-19 Domestic 0.447 14.0% 0.68 6,102 0.161

02/1946 to 07/1946 R-10 African 1.157 3.5% 0.44 8,492 0.145

06/1946 to 07/1946 Q-20 African 1.157 17.7% 2.21 82 0.007

Sum: 25,666 0.84

a. Source: Aerospace Corporation (1981), Table B-1. MT = metric ton (1,000 kg).

b. Per Table A-2. nt = nuclear transformation.

c. (Electron Energy Released per Decay)*(Average Grade) ÷ 8%.

d. The value for each ore ID is (Mass Processed)*(Average Beta Dose Rate Relative to 8% L-30)/(Total Mass Processed). 

The sum of the values is the approximate ratio of the average 1943 to 1946 ore beta dose to the ore dose due to 

8% L-30 ore.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 43 of 102

Table 4-6. Assigned beta dose rates, 1943 to 1946.

Category/job 

Median dose rate (rem/yr)

Hands and forearms Remainder of body GSD

High Variesa 74

Ball Mill Operator 221 74 1.52

Step I Process Operator 221 74 1.52

Weighmaster 221 74 1.52

Step II Process Operator 158 53 1.52

Loader 26 26 1.52

Medium 17.6 5.85

Chemist/Lab Technician 17.6 5.85 2.65

Engineer 17.6 5.85 2.65

Janitor 17.6 5.85 2.65

Laundry Worker 17.6 5.85 2.65

Maintenance Worker 17.6 5.85 2.65

Ore Sampler 17.6 5.85 2.65

Seamster, Seamstress 17.6 5.85 2.65

Step III Process Operator 17.6 5.85 2.65

Tool Crib Worker 17.6 5.85 2.65

Low 3.00 3.00

Draftsman 3.00 3.00 2.65

Fire Inspector 3.00 3.00 2.65

Guard 3.00 3.00 2.65

Nickel Operator 3.00 3.00 2.65

Nurse 3.00 3.00 2.65

Office Worker 3.00 3.00 2.65

Plant Superintendent, Asst. Supt. 3.00 3.00 2.65

Shipping and Receiving Clerk 3.00 3.00 2.65

Storekeeper 3.00 3.00 2.65

Tank Farm Operator 3.00 3.00 2.65

a. See Section 4.1.2.1.3.

Workers with jobs analogous to those assigned to the low beta exposure category for 1947 to 1949 

were assigned a beta dose rate of 3 rem/yr to the whole body. Table 4-6 shows the assignments.

1943 to 1946 Workers Ore Sampler

The ore sampler was considered to have an exposure potential similar to that of a chemist and was 

assigned dose rates of 17.6 rem/yr to the hands and forearms and 5.85 rem/yr to the remainder of the 

body.

4.1.2.1.2 Outdoor Dose Rate

The average outdoor beta dose rate to which workers would have been exposed during production 

was assumed to be at most equal to the indoor level in Building 30 based on the reasoning in 

Section 4.1.1. Therefore, the outdoor beta rate was estimated as 0.676 mrem/hr (based on beta dose 

rate at 3 ft before vacuum cleaning and flushing in Table 4-1). For 0.5 hr/wd exposure, 6 wd/wk, and 

50 wk/yr, the average worker exposure would have been 0.1 rem/yr. This is negligible in comparison 

with the indoor doses, given the approximate nature of the estimates, and was ignored.

4.1.2.1.3 Categories

Table 4-6 summarizes assigned beta dose rates for the 1943 to 1946 production period. All are 

considered to be median values of a lognormal distribution. The dose rates that were derived from 

the 1943 to 1946 data in Table 4-4 were judged to be based on near-maximum exposure parameters. 

They were assigned a GSD of 1.52, which corresponds to a distribution in which the ratio of 95th-
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 44 of 102

percentile value to median is equal to a factor of 2. The other dose rates were considered to be more 

uncertain. They were assigned a GSD of 2.65, which corresponds to a distribution in which the ratio 

of 95th-percentile value to median is equal to a factor of 5. 

To simplify dose reconstruction and take into account the uncertainties of the estimates, jobs were 

grouped into three categories – high, medium, and low – based on beta dose to the “remainder of the 

body” (all parts of the body except the hands and forearms). For all jobs in each category, the beta 

dose to the remainder of the body was assumed to be the highest value among the jobs in the 

category. 

Hands and Forearms Guidelines

If beta dose to the hands and forearms is needed for a particular dose reconstruction, it should be 

obtained as follows:

• For the low and medium categories, the values in Table 4-6 should be used.

• For the ball mill operator, Step I process operator, and weighmaster, the value in Table 4-6

should be used.

• For the Step II process operator, the dose to the hands and forearms should be taken as 

3 times the remainder-of-the-body dose for the high category or 221 rem.

• For the Loader, the dose to the hands and forearms should be taken as equal to the 

remainder-of-the-body dose for the high category or 74 rem.

4.1.2.2 Gamma

Gamma film badge data are available for Step I process workers for a portion of the 1943 to 1946 

production period. The data covers the period from January 31, 1944, through February 26, 1945,

except for a 3-week gap from April 18 to May 5, 1944, for which data were not obtained due to a film 

handling error (Ferry 1944d). Badges were usually worn for about a week, but during the first 

3 months some of the badging periods were longer (2 to 4 weeks). In the data analysis, each

multiweek measurement was treated as a set of individual weekly results with each equal to the 

weekly average that was determined from the multiweek measurement. For example, a 4-week result 

of 300 mR was treated as four individual 1-week results of 75 mR.

About 20 workers were typically badged in each period – about 10 to 12 Loaders and 7 to 9 

processors. It is not known whether all Step I personnel were badged. Film badges were provided by 

the Medical Section of the MED (presumably the University of Rochester).

From January 31, 1944, through March 27, 1944, results were in units of roentgen per 8-hour day (the 

designations "Roentgen," "R," and "r" in different reports of the data were interpreted as synonymous). 

After March 27, 1944, results were reported in tolerance units (designated "fractions of tolerance" or 

"times tolerance"). The MED tolerance value for gamma radiation in this period was stated as 

0.1 R/8-hr day and also as 0.5 R/wk and 0.6 R/wk (Ferry 1944e; Tybout 1945c). Linde employees are

reported to have worked a 6-day 48-hour workweek (Dupree1983a,b). For analysis of the gamma 

data, a 6-day workweek was assumed. To obtain roentgen per week, results in roentgen per 8-hour 

day were multiplied by 6 and results in tolerance units were multiplied by 0.6.

Some of the gamma results were reported in the form "<X," where the value of X varied. The average 

of the X values was 44 mR/wk. In the analysis of the data, 44 mR/wk was taken as the LOD, and 

results reported as "<X" or zero were considered to be <LOD results. However, all nonzero results, 

even if below 44 mR/wk, were treated as ≥LOD results.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 45 of 102

During the first part of the measuring period, the plant processed radium-containing L-30 African 

pitchblende ore. In an analysis of plant operations (Aerospace Corporation 1981, Table B-1), it was 

estimated that processing of the African ore began in December 1943 and continued through 

November 1944 and that processing of domestic ore with minimal radium content began in December 

1944. The film badge data showed a sharp drop in the number of ≥LOD results during the first 

2 weeks of November with the low point during the week ending November 12, so this was taken as 

the date when processing of the African ore ended. Exposures to workers were averaged separately 

for the periods from January 31 to November 12, 1944 (41 weeks) and November 13, 1944 to 

February 26, 1945 (15 weeks).

The data identified each worker's job activity. The job activities were grouped into categories. Each 

category consisted of jobs that were judged to have similar exposure potential based on job 

descriptions and the data. Table 4-7 lists the job activities and the category to which each was 

assigned.

African Ore

Table 4-8 displays statistical characteristics of the film badge data for the period from January 31, 

1944 to November 12, 1944, when African ore was processed. For each category, a lognormal 

distribution was assumed and the following parameters were determined:

• The number of film badge measurements,

• fD = dosimeter fraction = the fraction of results in a year at or above the LOD,

• fM = missed fraction = the fraction of results in a year below the LOD,

• mD = GM of all results above the LOD, and

• The GSD of all results above the LOD based on the assumption that their distribution was 

lognormal.

Table 4-7. Job categories for Step I gamma dose 

analysis, 1943 to 1946.

Job activitya Category

Ball mill Ball mill operator

Ball mill operator Ball mill operator

Cleaning up plant Cleanup

Loader Loader

Loader (handling ore) Loader

Loader foreman Loader

Ore sampler Ore sampler

Sampler Ore sampler

Barreling of tails from Moore filter Process operator

Chief operator Process operator

Digest Process operator

Digest & V.P. Process operator

Eimco & prod. precipitation Process operator

Foreman Process operator

Lead removal Process operator

Making ore digest in Pachuca tanks Process operator

Moore filter operator Process operator

Moore operator Process operator

Moore tailings Process operator

Moore tailings/Moore operator Process operator
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 46 of 102

Job activitya Category

Moores Process operator

Pachuca digest Process operator

Pachuca tanks Process operator

Product precipitation Process operator

V.P. press Process operator

V.P. removal Process operator

Weighmaster Weighmaster

a. Job activity descriptions are from the film badge records.

Table 4-8. Step I gamma exposure data for African ore processing, January 31 to November 

12, 1944.

Category

#

Weekly

results

Dosimeter

fraction

fD

Missed

fraction

fM

For results in dosimeter fraction

GM mD

(mR/wk) GSD

Ball mill operator 17 0.9412 0.0588 93 1.81

Loader 507 0.8383 0.1617 123 2.00

Ore sampler 36 0.5000 0.5000 109 2.61

Process operator 202 0.6980 0.3020 71 1.93

Weighmaster 31 0.8387 0.1613 100 2.15

The subscripts D and M denote dosimeter dose and missed dose, respectively, as defined by NIOSH 

(2007b, pp. 30 and 32). For the current set of data, the value of fD was estimated as the number of 

≥LOD results divided by the number of measurements, and fM was estimated as the number of <LOD 

results divided by the number of measurements. (Case-by-case consideration is required when the 

number of measurements is very low. Such situations are dealt with in Sections 4.1.4.1 and 4.1.4.2.)

The estimated median annual dosimeter dose DD for a 50-week year is shown in Table 4-9. This was 

calculated as follows:

DD = 50fDmD (4-1)

Table 4-9. Step I gamma exposure rate parameters for 

African ore processing, January 31 to November 12, 1944.

Category

DD

a

(R/yr) GSDb

DM

a

(R/yr) GSD DD + DM

Ball mill operator 4.37 2.61 0.06 1.52 4.44

Loader 5.17 2.61 0.18 1.52 5.35

Ore sampler 2.72 2.61 0.55 1.52 3.27

Process operator 2.47 2.61 0.33 1.52 2.80

Weighmaster 4.21 2.61 0.18 1.52 4.39

a. Each dose value represents the median of a lognormal distribution.

b. Highest GSD for DD values in Table 4-8 is assumed to apply to all 

categories.

For simplicity and because a GSD value based on small number of measurements might not be valid, 

the highest GSD in Table 4-8 was assigned to every job category. 

In accordance with NIOSH (2007b), the annual missed dose was assumed to have a lognormal 

distribution with median DM and the 95% confidence limit DM95. For a 50-week year, these were 

calculated as follows:

DM = 50fM(LOD/2) (4-2)

DM95 = 50fM(LOD) (4-3)
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 47 of 102

A GSD of 1.52 was calculated from DM and DM95. The results for DM are in Table 4-9.

Domestic Ore

The analysis of exposure data for the period November 13, 1944, to February 26, 1945, when 

domestic ore with minimal radium content was being processed, was similar to that made for the 

preceding period. The results are displayed in Table 4-10 and Table 4-11. For routine processing of 

domestic ore, the weighmaster and ball mill operator were assigned the highest annual dose. Most of 

this dose is missed dose.

Table 4-10. Step I gamma exposure data for cleanup and domestic ore processing, 

November 13, 1944, to February 26, 1945.

Category

#

Weekly

results

Dosimeter

fraction

fD

Missed

fraction

fM

For results in dosimeter fraction

GM mD

(mR/wk) GSD

Ball Mill Operator 0 (a) (a) (a) (a)

Cleanupb 15 0.3333 0.6667 30.0 1.00

Loader 164 0.0061 0.9939 34.2 (c)

Ore Sampler 11 0.0000 1.0000 0.0 (d)

Process Operator 74 0.1486 0.8514 32.9 1.17

Weighmaster 11 0.1818 0.8182 35.5 1.27

a. Not available because no weekly data was reported.

b. Cleanup was an occasional short-term activity that was not part of routine domestic ore processing.

c. Not available because only one result was ≥LOD.

d. Not available because no results were ≥LOD.

The results obtained for this period also apply to the processing of scrap and residues from uranium 

processing. The uranium processing would have removed most of the content of radium and radium 

progeny.

Table 4-11. Step I gamma exposure rate parameters for cleanup 

and domestic ore processing, November 13, 1944, to February 

26, 1945.

Category

DD

a

(R/yr) GSD

DM

a

(R/yr) GSD DD + DM

Relative

DD + DM

b

Ball mill operatorc 0.32 1.27 0.90 1.52 1.22 28%

Cleanupd 0.50 1.27 0.73 1.52 1.23 (e)

Loader 0.01 1.27 1.09 1.52 1.10 21%

Ore sampler 0.00 1.27 1.10 1.52 1.10 34%

Process operator 0.24 1.27 0.94 1.52 1.18 42%

Weighmaster 0.32 1.27 0.90 1.52 1.22 28%

a. Each dose value represents the median of a lognormal distribution.

b. DD + DM for domestic ore processing as percent of its value for same 

category for African ore processing.

c. There were no measured data available for the ball mill operator. This 

category was assigned the same dose as weighmaster because of the 

similarity of their exposure potential and because of the possibility that 

processing of some types of domestic ore might require handling by the Ball 

Mill Operator.

d. Cleanup was an occasional short-term activity that was not part of routine 

domestic ore processing.

e. Not available.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 48 of 102

4.1.2.2.1 Estimation of Worker Doses

African Ore

The annual gamma exposures of the monitored Step I workers (Table 4-9) were considered to be 

sufficiently similar that all could be assigned the same value. For dose reconstruction purposes, all 

were assigned an exposure rate of 5.35 R/yr, the highest value of DD + DM in Table 4-9, and a GSD of 

2.61, the highest GSD in Table 4-9.

Step I Process Workers

For other workers, most of the task-related sources of gamma exposure were probably comparable to 

or weaker in effect than they were for Step I process workers. Step II and Step III process workers 

were handling refined uranium materials with much lower gamma emission rates because much of the 

radium and radium progeny had been removed. Nickel process workers were handling 

nonradioactive materials. In general, support personnel – such as laboratory, maintenance and 

janitorial personnel – had less frequent exposure to radium-containing materials or dealt with smaller 

quantities or worked at greater distances from the materials. They also worked in less dusty 

environments and so had less exposure to airborne radioactivity and to floor and wall contamination. 

Other Workers

One source of gamma exposure, however, was more dependent on worker location than on worker 

task. This was radiation from large quantities of ore that were stored in ore piles. A report of radiation 

surveys on March 2 and 3, 1944 (Ferry 1944f) identified two 200,000-lb piles of 10% ore and one 

300,000-lb pile of 10% ore "in the receiving room" as well as one 30- by 30- by 12-ft pile of 8% ore at 

an unspecified location (percentages specify weight % of U3O8). Based on an analysis of typical 

pitchblende ore at Linde, the ore density was estimated as 3.25 g/cm3 (Aerospace Corporation 1981, 

Table C-1). Therefore, the 30- by 30- by 12-ft pile contained approximately 2 million lb of ore and was 

the largest of the four piles.

The probable location of the ore was the shipping and receiving platform at the south end of 

Building 30 (Figure 2-3). This would have placed the ore near the "grizzly enclosure" through which 

ore was dumped onto the conveyor belt at the start of Step I processing. A letter to Linde reporting on 

a March 2, 1944, inspection recommended cleaning up and rearranging the platform so that ore would 

be stored in its southeast corner (Cranch 1944a), and a letter from the Ceramics Plant on May 8, 

1944, indicated compliance with this recommendation (Rehm 1944b).

In Ferry (1944f), an exposure rate of 0.23 R/8-hr day was reported at 1.5 ft from the 30- by 30- by 

12-ft pile. Potential exposures at more distant locations were estimated by mathematical analysis. 

For the region very close to an ore pile, the dependence of exposure rate on distance was determined 

by fitting data from Linde pilot plant studies for a 4- by 5- by 20-ft pile of 8% ore (Skinner 1944; 

Wiesendanger 1944). The fit was obtained using the MicroShield computer program (MicroShield 

Team 2003), which is based on the point kernel methodology. The values from the program were 

multiplied by a single scaling factor that was chosen to provide a best fit to the measured data. The 

adjustment compensated for simplifications in the modeling. Figure 4-1 shows the data and the fit. 

Gamma exposure rates measured during pilot plant studies at Linde Ceramics (Skinner 1944) are 

shown by the diamond symbols. The rectangles show the fit to the data with distance assumed to be 

measured along the perpendicular bisector of a 4- by 20-ft. face. The ore grade (8%) was assumed to 

be that determined in the analyses reported by Wiesendanger (1944, p. 18).The data covers the 

range of 0 to 10 ft from the 4- by 20-ft face. The adjusted MicroShield output fit the data to within 

±10% between 3 ft and 10 ft from the face. The results were extended to larger distances and to a 

30- by 30- by 12-ft pile by more MicroShield calculations that were scaled with the same factor. 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 49 of 102

Figure 4-1. Fit to gamma exposure data for a 4- by 5- by 20-ft 

pile of 8% African pitchblende ore. 

Figure 4-2 shows the MicroShield results and an analytic function fit to them. The diamond-shaped 

symbols show calculated values of exposure rate as a function of distance from the pile center. 

Distance is measured along the perpendicular bisector of a 12- by 30-ft face. The values were 

calculated using MicroShield (MicroShield Team 2003). The line is a plot of the analytic function 

obtained by a least-squares fit to the MicroShield results. The function is:

(1/r

2

) × exp[a × exp(–b × r) + (c–d × r)] (4-4)

where a = 1.906, b = 3.126×10-02, c = 7.660, and d = 1.521×10-03. The fit agreed with the calculated 

values to within ±3.0% over the range of 16 to 2,000 ft from the center of the pile.

Exposure Rates Near 4 ft x 5 ft x 20 ft Ore Pile

0

5

10

15

20

25

30

35

40

45

0 2 4 6 8 10

Distance from 4x20 Face (ft)

Exposure Rate (mR/h)

Data

Fit
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 50 of 102

Calculated Exposure Rate 

Due to 30 ft x 30 ft x 12 ft Ore Pile

1.000E-05

1.000E-04

1.000E-03

1.000E-02

1.000E-01

1.000E+00

1.000E+01

1.000E+02

1 10 100 1,000 10,000

Distance from Pile Center (ft)

mR/h

Data

Fit

Figure 4-2. Calculated exposure rates and analytic fit for a 30- by 30-

by 12-ft pile of 8% African pitchblende ore. 

The function facilitates calculation of exposures at arbitrary distances. As a check on the 

methodology, the analytic function was used to calculate the exposure rate at a point where it had 

been measured, which was 1.5 ft from the face. The result was 0.19 R/8 hr, which is 17% below the 

measured value. This degree of difference is consistent with the uncertainties typical of this type 

measurement and the approximations inherent in the MicroShield code. 

The calculated results are expected to overestimate radiation levels well away from the pile. The 

calculations account for air attenuation, but not for attenuation due to objects such as walls, floors, 

and equipment. In addition, the calculations are based on distance along the perpendicular bisector 

of a pile face. Exposure rate at any distance is lower for a point not on the perpendicular bisector.

To estimate the impact on the other workers in Building 30, who did not wear dosimeters, exposure 

rates due to a 30- by 30- by 12-ft pile in the southeast corner of Building 30 were calculated for 

various facilities in the building. In each case, the exposure rate was calculated for the point closest 

to the center of the pile. The calculated rates are in Table 4-12. They range from 0.28 R/yr (for the 

laboratory) to 4.9 R/yr (for the lunchroom). The results indicate a potential for exposures to the other 

workers in Building 30 comparable to the exposures to Step I process workers.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 51 of 102

Table 4-12. Estimated maximum exposure rates in 

selected Building 30 facilities due to 30-by 30- by 12-

ft pile of 8% ore in southeast corner. 

Facility

Distance

(ft)a

Exposure rate

(R/yr)

b

Bag washer room 73 1.09E+00

Laboratory 128 2.81E–01

Locker & wash rooms 43 4.46E+00

Lunchroom 42 4.89E+00

Sampling room 57 2.10E+00

a. Distance is the distance from the pile center to the closest 

point that is inside the facility and 3 ft above the floor.

b. Based on 2,550 hr/yr exposure.

Although few of the other workers spent the major portion of their work time near the southeast corner 

of Building 30, some might have worked near other ore piles, near tailings piles, or near large 

quantities of ore in process. Because the locations and quantities of these other sources are not 

known in sufficient detail, any worker who spent a substantial amount of time in Building 30 had a 

potential for significant gamma exposure. Such workers included not only the production personnel in 

the building, but also nurses (whose office was there), maintenance personnel (who came there to 

maintain, repair, or refurbish), janitors, personnel who used locker rooms or the lunchroom, and office 

personnel who might have had to regularly visit the building (e.g., product accountants, industrial 

relations representatives, the fire marshal, and engineers). Last, it appears that there were office 

personnel who were permanently stationed in the building. A February 1944 letter from the plant 

addressing the frequency of required medical examinations states that it would not be appropriate to 

reduce the examination frequency for all office personnel because “we have some stenographers and 

clerks, stores clerks, etc., who work in the main building and are from time to time exposed to the 

same hazards as the operating personnel” (Holmes 1944a). In view of the likely difficulty of 

establishing that a particular worker did not spend a substantial portion of time in Building 30, for dose 

reconstruction it is assumed that all plant personnel had a significant potential for gamma exposure. 

All Ceramics Plant employees were assigned a gamma dose of 5.35 R/yr for the whole body, based 

on the results for the most exposed group of monitored process workers (Table 4-9).

Domestic Ore

During domestic ore processing, the gamma exposure rates were lower because the ore contained 

lower proportions of radium and other uranium series progeny than African ore. However, it is difficult 

to estimate how much lower because of uncertainty as to how much African ore radioactivity remained 

in the preprocessed ores at Linde (Section 2.2). The value of DD + DM for Step I process operators 

during domestic ore processing (Table 4-11) is 42% of the value during African ore processing (Table 

4-9). This difference is judged not to be large enough to justify the complication of using lower doses 

for periods of domestic ore processing. Therefore, for dose reconstruction during the 1943 to 1946 

production period, the gamma exposure rate of 5.35 R/yr was assumed to apply for the whole period.

4.1.2.2.2 Outdoor Exposure Rate

The average outdoor gamma exposure rate to which workers could have been exposed during 

production was assumed to be at most equal to the indoor level in Building 30 based on the reasoning 

in Section 4.1.1. Therefore, the outdoor gamma rate was estimated as 0.131 mR/hr based on the 

level before vacuuming and flushing in Table 4-1. For 0.5 hr/wd exposure, 6 wd/wk, and 50 wk/yr, the 

average worker exposure would have been 0.020 R/yr. This is negligible in comparison with the 

indoor exposures given the approximate nature of the estimates and was ignored.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 52 of 102

4.1.3

Little information is available about the status of activities during the standby period. It is likely that 

the onsite staff consisted primarily of a small number of management and janitorial personnel – both 

of whom worked primarily in an office environment – and guards. For dose reconstruction, each 

worker during standby was classified as either a guard or a general worker, and worker time was 

assumed to have been spent in an office building, in production buildings, and outdoors. Averaged 

over the entire standby period, each worker's allocation of time was assumed to have been as 

indicated by the occupancy factors in Table 4-13.

Standby, 1946 to 1947

Table 4-13. Ceramics Plant beta and gamma radiation rates during standby.

Parameter

Category Time-weighted radiation ratea

Office Production Outdoors Beta (rem/yr) Gamma (R/yr)

Beta (mrem/hr) 0.000 0.676 0.676 (b) (b)

Gamma (mR/hr) 0.000 0.131 0.131 (b) (b)

Occupancy factor

General worker 0.833 0.111 0.056 3.04E-01 5.91E-02

Guard 0.756 0.111 0.133 4.46E-01 8.67E-02

a. Based on 9.0 hr/d exposure, 6 d/wk, 50 wk/yr. Based on the underlying data and judgment, a GSD of 

3 is assigned. The beta and gamma rates are for the whole body.

b. Not applicable.

Gamma radiation levels were measured at six locations in Building 30 on October 22, 1946. 

Measurements were made at 1 in. from the surface of interest. The results were reported as 0 R/8 hr

for four of the locations and 0.005 R/8 hr (0.625 mR/hr) for the other two locations (each near an ore 

dumping grill) (Howland 1946). Because the dumping grill was one of the most contaminated spots in 

the plant, the exposure rate there was not considered typical of the conditions that would have been 

encountered upon occasional entry during standby. Instead, the indoor gamma and beta levels for a 

production building were taken as the values in Table 4-1 before vacuum cleaning and flushing. 

Outdoor gamma and beta levels were taken as equal to the indoor rates based on the reasoning used 

above in the discussion of the preproduction period. The gamma and beta radiation rates in an office 

building were assumed to be zero.

Table 4-13 summarizes the calculation of annual radiation rates based on the above parameters.

Because there would have been little need for direct handling of radioactive materials by Ceramics 

Plant workers in this period, beta dose rate to the hands and forearms was taken as equal to the beta 

dose rate to the remainder of the body.

4.1.4

4.1.4.1 Beta

Rehabilitation and Production, 1947 to 1949

For part of the period from 1947 to 1949, weekly film badge measurements of beta and gamma 

exposure were available (LAPC undated b). The data had been "de-identified" by removal of 

personnel names (Wallace 2003), but many of the records were labeled with a job title. The data file 

contained about 6,000 records from January 7, 1948 through December 12, 1949. By comparison 

with a copy of a laboratory film badge report (AEC 1949b), it was determined that the date identified 

the beginning of the week when the badge was worn. For records before March 15, 1948, badges 

were processed by the University of Rochester (Heatherton 1948g, Osinski 1948a); thereafter, they 

were processed by the AEC New York Directed Operations Office Radiological Laboratory (Osinski 

1948b; AEC 1949b). When use of the badges was initiated, the intention was for them to "be 

available for use by all entering a contaminated area" (Heatherton 1947).
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 53 of 102

The computer file had three columns of data but no labels to indicate units or type of radiation. The 

third column was obviously the total radiation dose. The first column was determined to be beta 

radiation and the second gamma radiation by comparison with a laboratory film badge report (AEC 

1949b). The report stated both the beta and gamma units to be "exposure, mr." Progress reports by 

the Linde health physicist referred to the film badge results sometimes as "mr" and sometimes as 

"mrep" (Heatherton 1949d,e). The beta units are designated here as mrem and are assumed for 

purposes of dose reconstruction to be equivalent to shallow dose at 0.07 mm, Hp(0.07). Before 

March 15, 1948, a beta value below the limit of detection was reported as "<80." After March 15, 

1948, the limit of beta detection was not explicitly stated, but it was deduced to be 35 mrem from an 

examination of the data. Some of the results were reported as "0", "*" or "**". These were interpreted 

as meaning less than the LOD. 

The beta film badge data was analyzed to determine dose rates for various job categories. 

Approximately one-third of the badge records were labeled with the job title NA. These were ignored 

because it could not be determined which had been worn by a worker and which had been used for 

other purposes (e.g., as controls or to measure process dose). Job titles that were specified in the 

records were binned into categories that combined jobs that were judged to have had similar 

exposure potential. Table 4-14 displays the job titles in each category. 

Table 4-14. Job categories for 1947 to 1949 Step III film badge analysis.

Job title Category Job title Category

AEC Office Maint g l Maintenance

Asst engineer Engineer Maint group leader Maintenance

Asst supt Superintendent Maintenance Maintenance

Carpenter Maintenance Millwright A Maintenance

Chem oper Process operator Millwright C Maintenance

Chem oper A Process operator Moveman Loader

Chem oper B Process operator Painter Maintenance

Chem oper C Process operator Painter B Maintenance

Chem oper G L Process operator Personnel Office

Chemist Chemist Pipefitter Maintenance

Control engr Engineer Process foreman Process operator

Electrician Maintenance Prod accounting Office

Electrician hlpr Maintenance Property dept Office

Fire chief Engineer Safety engr Engineer

Foreman (unspecified type) (a) Seamster Seamster

Group leader (unspecified type) (a) Security agent Office

Health physicist Engineer Shift storekpr Storekeeper

Janitor Janitor Shift stores att Storekeeper

Labor foremanb Loader Shipping & rec Storekeeper

Laundry helper Laundry worker Storekeeper Storekeeper

Laundry worker Laundry worker Superintendent Superintendent

Loader Loader Tool crib keeper Storekeeper

Loader foreman Loader Trades helper Maintenance

Mail girl Office Truck driver Truck driver

Maint Maintenance Weighmaster Weighmaster

Maint foreman Maintenance Welder Maintenance

a. There were two records with the job title Foreman and one with the job title Group Leader. They were not included in the gamma dose 

analysis because the title was not descriptive enough to identify the work activity.

b. From available data (Heatherton 1949b; AEC 1949b), it was deduced that the worker with the job title Labor Foreman was supervising 

loaders.

Table 4-15 displays statistical characteristics of the beta film badge data. The data were assumed to 

have a lognormal distribution. The parameters fD, fM, and mD were defined as in the analysis of the 

1943 to 1946 gamma dosimetry data (Section 4.1.2.2). For categories in which there were 50 or more 

badge results, fD was taken as equal to the number of results at or above the LOD divided by the total 

number of badge measurements and fM was taken as (1 – fD). Categories with fewer than 50 results 

were examined on a case-by-case basis. If the workers in the category were rarely badged, it was 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 54 of 102

assumed that they only occasionally entered the Linde radiation area. The value assigned to fD was 

the number of results at or above the LOD divided by 50, which is the number of workweeks per year. 

The missed fraction fM was taken as (1/50) times the highest number of results below the LOD for any 

member of the category. In the present case, this alternative method was applied to the categories 

Office and Superintendent.

Table 4-15. Step III beta dose data, 1947 to 1949.

Job

#

Badges

Dosimeter 

fraction

fD

Missed 

fraction

fM

For results ≥LOD

GM mD

(mrem/wk) GSD

Chemist 183 0.0055 0.9945 95 (a)

Engineer 243 0.0535 0.9465 113 2.12

Janitor 261 0.0421 0.9579 132 2.17

Laundry 115 0.1565 0.8435 106 1.65

Loader 304 0.0559 0.9441 78 1.47

Maintenance 913 0.0526 0.9474 146 1.62

Office 9 0.0000 0.0600 (b) (b)

Process operator 1,814 0.1527 0.8473 145 1.81

Seamster 44 0.0909 0.9091 141 1.31

Storekeeper 207 0.0048 0.9952 120 (a)

Superintendent 3 0.0000 0.0400 (b) (b)

Truck driver 18 0.0000 1.0000 (b) (b)

Weighmaster 61 0.0492 0.9508 136 1.42

a. Only one result was ≥LOD; GSD could not be determined.

b. No results ≥LOD; GM and GSD could not be assigned.

The estimated median annual dosimeter dose DD and the median annual missed dose DM for a 

50-week year and their GSDs were determined in the same way as for the 1943 to 1946 gamma data 

(Equations 4-1, 4-2, and 4-3 in Section 4.1.2.2). Because badges would not have been worn until 

workers arrived indoors, an outdoor beta dose rate of 0.10 rem/yr (from Section 4.1.2.1.2) was added 

to the DM values. For the office worker and superintendent, the outdoor contribution was larger than 

the indoor DM value, so the outdoor GSD of 3 was assigned. For the other categories, the outdoor 

contribution was not the major portion of the sum, so the GSD was taken as not changed by the 

addition. Table 4-16 displays the results.

For most jobs, the largest contribution to DD + DM + OD was missed dose DM. The production workers 

in 1947 to 1949 were probably scattered among three buildings – Building 38, the location of the Step 

III processing facilities; Building 30, which contained a loading dock, locker rooms, lunchroom, nurses’ 

office, and chemistry laboratory; and Building 31, which contained shop facilities. None of the 

buildings had been decontaminated. The average beta radiation level in Building 30 for this period 

was estimated (Section 4.1.1) as 0.676 mrem/hr (value before vacuuming and flushing in Table 4-1), 

which is equivalent to 34.5 mrem/wk and 1.7 rem/yr for a person spending 8.5 hr/d, 6 d/wk, and 50 

wk/yr in the building. This is comparable to the film badge LOD for beta radiation (35 mrem/wk) and 

so might have been a substantial portion of the missed dose. Based on this estimate of 1.7 rem/yr of 

possible dose and the lowest film badge DD + DM value in Table 4-16 of approximately 1 rem/yr for 

workers, who were regularly badged, it was assumed that all workers, even office workers, received a 

beta dose of at least 1 rem/yr. Office workers were included because it is unknown which office 

workers spent substantial portions of their time in production buildings. Some workers who might 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 55 of 102

Table 4-16. Beta dose rates for Step III, 1947 to 1949.

Job

DD

a DM + ODa DD + DM

+OD

(rem/yr) GSD (rem/yr) b (rem/yr) GSD

Process operator 1.11 2.17 0.84 1.52 1.95

Laundry 0.83 2.17 0.84 1.52 1.67

Seamster 0.64 2.17 0.90 1.52 1.54

Maintenance 0.38 2.17 0.93 1.52 1.31

Weighmaster 0.33 2.17 0.93 1.52 1.26

Engineer 0.30 2.17 0.93 1.52 1.23

Janitor 0.28 2.17 0.94 1.52 1.22

Loader 0.22 2.17 0.93 1.52 1.15

Chemist 0.03 2.17 0.97 1.52 1.00

Storekeeper 0.03 2.17 0.97 1.52 1.00

Truck Driver (c) (c) 0.98 1.52 0.98

Office (c) (c) 0.15 3.00 0.15

Superintendent (c) (c) 0.14 3.00 0.14

a. Each DD and DM value represents the estimated median of a lognormal 

distribution. OD = estimated outdoor dose rate.

b. The highest GSD from Table 4-15 was assumed for all categories for which a 

GSD can be assigned.

c. No result ≥LOD. 

have had offices in the office building might also have spent much of their workday in the production 

area (e.g., engineers, health physicist, and fire marshal), and some clerical workers might have 

actually had work locations in the production area [discussion near the end of Section 4.1.3.2.1 and 

Holmes (1944a)].

Categories

For dose reconstruction, the jobs were grouped into two categories – medium and low. The medium 

category was assigned a dose rate of 1.95 rem/yr. Workers whose jobs might have required frequent 

handling of radioactive materials were placed in the medium category. The tool crib attendant was 

assigned to the medium category because of the possible handling of contaminated tools. The truck 

driver was placed in the medium category because it was the practice at the plant in 1943 to 1946 for 

truck drivers to work as loaders when not driving a truck (LAPC 1945b, p. 10); it was assumed that the 

same practice applied from 1947 to 1949. Engineers and the plant health physicist (considered a type 

of engineer) were placed in the medium category because of their roles in closely monitoring and 

troubleshooting production activities. 

The low category was assigned a dose rate of 1.00 rem/yr. Assigned to it were most office workers, 

the superintendent, and storekeepers. It also included guards and the fire marshal. 

A list of jobs with the category assignment of each is provided in Section 4.4.

Rehabilitation

Beta dose rates during rehabilitation were assumed to be the same as during production to account 

for exposures while cleaning contaminated facilities, setting up, and testing equipment.

Hands and Forearms Dose

Film badges were worn on the chest (Heatherton 1948h). Measurements during various Step III 

process operations showed that dose rates to other areas of the body differed from dose rates at the 

film badge location with the difference dependent on the operation and the area of the body 

(Heatherton 1948h). The results are shown in Table A-6. Based on these data, it was estimated for 

this site profile that the average dose to the hands and forearms of a worker handling radioactive 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 56 of 102

material was 3 times the chest dose and that average dose to the remainder of the body was equal to 

the chest dose. For each body area, the estimate was based on the average ratio for the studied 

operations with the result rounded up to the nearest integer. No credit was taken for wearing gloves. 

For dose reconstruction, a hands and forearms dose of 5.85 rem/yr was assumed for the medium 

category. Workers in the low category were assumed to have a hands and forearms dose of 1 rem/yr, 

which is the same as the dose assigned to the remainder of the body.

4.1.4.2 Gamma

As discussed in Section 4.1.4.1, weekly film badge measurements of beta and gamma exposure 

during 1947 to 1949 production were available (LAPC undated b). As noted, the measurement units 

were specified as "mr" or "mrep." It is assumed here that for gamma radiation the units were 

equivalent to milliroentgen. Results reported as "0", "*", or "**" were interpreted as meaning less than 

the LOD. For records before March 15, 1948, a gamma value below the LOD was reported as "0" or 

"<50." For records after March 15, 1948, the limit of gamma detection was not explicitly stated but 

was deduced to be 45 mR from an examination of the data. A few scattered reports of nonzero 

exposures below 45 mR were judged not to reflect routine measuring capabilities.

The procedure for analyzing the gamma data and determining annual gamma exposures was the 

same as that for the analysis of the beta data (Section 4.1.4.1). Table 4-17 summarizes the statistical 

characteristics of the gamma data.

Table 4-17. Step III gamma exposure data, 1947 to 1949.

Category

#

Badges

Dosimeter

fraction

fD

a

Missed

fraction

fM

b

For results ≥LOD

GM mD

(mR/wk) GSD

Chemist 183 0.0164 0.9836 82 1.41

Engineer 243 0.0247 0.9753 145 2.13

Janitor 261 0.0115 0.9885 62 1.27

Laundry 115 0.0783 0.9217 81 1.46

Loader 304 0.0789 0.9211 141 2.14

Maintenance 913 0.0164 0.9836 87 1.38

Office 9 0.0200 0.0400 80 (c)

Process operator 1,814 0.0309 0.9691 86 1.52

Seamster 44 0.0000 1.0000 (d) (d)

Storekeeper 207 0.0048 0.9952 65 (c)

Superintendent 3 0.0200 0.0400 50 (c)

Truck driver 18 0.0000 1.0000 (d) (d)

Weighmaster 61 0.0000 1.0000 (d) (d)

a. fD = fraction of measurements ≥LOD.

b. fM = fraction of measurements <LOD.

c. Only one result was ≥LOD; GSD could not be determined.

d. No results ≥LOD: GM and GSD could not be assigned.

The estimated median annual dosimeter dose DD and median annual missed dose DM for a 50-week 

year and their GSDs were determined in the same way as for the 1947 to 1949 beta data 

(Section 4.1.4.1). Because badges would not have been worn until workers arrived indoors, an 

outdoor gamma exposure rate of 0.02 R/yr (Section 4.1.2.2.2) was added to the DM values. Because 

the outdoor contribution was not the major portion of any of the sums, the GSD was taken as not 

changed by the addition. Table 4-18 displays the gamma exposure rates that were determined.
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Table 4-18. Gamma exposure rates for Step III, 1947 to 1949.

Job

DD

a DM + ODa DD + DM + OD

(R/yr) GSD (R/yr) b (R/yr) GSD

Loader 0.56 2.14 1.06 1.52 1.61

Laundry 0.32 2.14 1.06 1.52 1.37

Engineer 0.18 2.14 1.12 1.52 1.30

Process operator 0.13 2.14 1.11 1.52 1.24

Maintenance 0.07 2.14 1.13 1.52 1.20

Chemist 0.07 2.14 1.13 1.52 1.19

Janitor 0.04 2.14 1.13 1.52 1.17

Storekeeper 0.02 2.14 1.14 1.52 1.15

Seamster (c) (c) 1.14 1.52 1.14

Truck driver (c) (c) 1.14 1.52 1.14

Weighmaster (c) (c) 1.14 1.52 1.14

Office 0.08 2.14 0.06 1.52 0.14

Superintendent 0.05 2.14 0.06 1.52 0.11

a. Each DD and DM value represents the estimated median of a lognormal 

distribution. OD = estimated outdoor dose rate.

b. The highest GSD from Table 4-17 was assumed for all categories for which a 

GSD can be assigned.

c. No results ≥LOD. 

As in the case of 1947 to 1949 beta dose (end of Section 4.1.4.1), it is to be expected that there was a 

contribution to gamma dose due to floor and wall contamination in the buildings. The average gamma 

radiation level in Building 30 for this period is estimated in Section 4.1.1 as 0.131 mR/hr (value before 

vacuuming and flushing in Table 4-1), which is equivalent to 6.7 mR/wk and 0.34 R/yr for a person 

who spent 8.5 hr/d, 6 d/wk, and 50 wk/yr in the building. This is below the film badge LOD for gamma 

radiation (45 mR/wk). Any clerical personnel who worked in production buildings (see Holmes 1944a) 

were assumed to have received this estimated floor and wall radiation plus the 0.14 R/yr that was

received by the occasionally badged office worker for a total of 0.48 R/yr. Because it might not be 

easy to identify which clerical personnel worked in a production building, all unbadged or infrequently 

badged personnel were assumed to have received 0.48 R/yr.

Categories

For dose reconstruction, the jobs were grouped into two categories – medium and low – with the 

gamma category assignment for each job the same as its 1947 to 1949 beta category assignment 

(Categories discussion in Section 4.1.4.1). Jobs in the medium gamma category were assigned an 

exposure rate of 1.61 R/yr for the whole body. Jobs in the low category were assigned an exposure 

rate of 0.48 R/yr for the whole body.

Rehabilitation

Gamma exposure rates during rehabilitation were assumed to be the same as during production in 

order to account for exposures while cleaning contaminated facilities, setting up, and testing 

equipment.

4.1.5

Cleanup of the Ceramics Plant is discussed in Section 2.6. Floor and wall radiation levels, which 

were measured at the start and end of the 1949 to 1950 decontamination of Building 30, were 

available and are summarized in Table 4-1. In addition, results were available from film badge 

measurements of beta and gamma radiation during the 1948 removal of equipment from Building 30. 

These results are summarized in Table 4-19. Because the results in Table 4-19 are film badge 

measurements, they include floor and wall radiation in addition to radiation from contaminated 

equipment.

Cleanup, 1949 to July 7, 1954
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Table 4-19. Beta and gamma radiation measured during 

equipment removal, 1948.

Beta dose rate

DD (rem/yr)

a GSD DM (rem/yr)a GSD

1.99 1.53 0.52 1.52

Gamma exposure rate

DD (R/yr)

a GSD DM (R/yr)a GSD

0.89 1.70 0.94 1.52

a. DD and DM are dosimeter dose and missed dose, 

respectively, as defined in NIOSH (2007b), pp. 30 and 32.

For external dose reconstruction, each worker is classified in one of the following categories:

• Cleanup worker, 

• Cleanup support worker, and 

• Non-cleanup worker.

A cleanup worker is defined as a worker who was directly engaged in the removal of radioactive 

equipment, components, or contamination (e.g., by sandblasting, vacuuming, or washing) for a 

substantial portion of the workday. This category includes equipment operators (such as a 

sandblaster operators) and laborers who were involved in moving and cleaning activities. A cleanup 

support worker is one who spent a substantial portion of the workday in the building or area being 

decontaminated to support the cleanup activities but was not actively engaged in the removal 

activities. Examples are the plant health physicist and a stores worker in a building being cleaned. All 

remaining plant personnel are considered non-cleanup workers. For a cleanup worker, the radiation 

estimate for both beta and gamma was taken as:

DD + DM + OD (4-5)

where DD and DM were taken from Table 4-19. The parameter OD is the outdoor dose rate. As for 

the preproduction and standby periods, it was estimated as equal to the estimated indoor floor and 

wall level at 3 ft before vacuum cleaning and flushing in Table 4-1. A cleanup support worker was 

assumed to have only half as much exposure to indoor radiation. The radiation level estimate was 

taken as:

0.5(DD + DM) + OD (4-6)

for both beta and gamma. Finally, a non-cleanup worker was assumed to have only 5% of the 

exposure to gamma radiation of a cleanup worker and to have a beta exposure equal to that 

calculated for Tonawanda Plant worker’s after 1946 (numbers were not adjusted for exposure time). 

The radiation level was estimated as:

0.05(DD + DM) + OD (4-7)

for gamma radiation. The beta radiation values specified are for all parts of the body. Table 4-20

presents the results. 

The beta dose rates in Table 4-20 are based primarily on film badge measurements. The film badges 

were worn on the chest (Heatherton 1948h). The hands and forearms would have been closer than 

the chest to radioactive materials during some of the steps involved in cleanup (e.g., dismantling and 

removing equipment, and scrubbing contaminated surfaces). For dose reconstruction, the beta dose 

rate to the hands and forearms of a cleanup worker was taken as 3 times that to the remainder of the 

body. The factor three was based on measurements discussed in Hands and Forearms Dose in 

Section 4.1.4.1.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 59 of 102

Table 4-20. External dose reconstruction parameters for cleanup.

Period/category

Betaa Gammab

Median (rem/yr)c GSDd Median (R/yr)c GSDd

07/01/1949–12/31/1950, 6-day week 

Cleanup worker 2.61E+00 4.04 1.85E+00 4.04

Cleanup support worker 1.36E+00 4.04 9.34E-01 4.04

Non-cleanup worker 3.26E-01 3.00 1.11E-01 4.04

01/01/1951–07/07/1954, 5-day week

Cleanup worker 2.18E+00 4.04 1.54E+00 4.04

Cleanup support worker 1.13E+00 4.04 7.78E-01 4.04

Non-cleanup worker 3.26E-02 3.00 9.26E-02 4.04

a. For the cleanup support and non-cleanup workers, the indicated beta rate is to the whole 

body. For the cleanup worker, the rate is to all parts of the body except the hands and 

forearms. For these, the rate is 3 times higher. See text.

b. Gamma exposure rates are for the whole body.

c. Annual rates based on 8.5 hr/wd indoors, 0.5 hr/wd outdoors, 50 wk/yr, except as noted for 

the non-cleanup worker.

d. The GSD for the beta value for the non-cleanup worker is the estimated GSD for outdoor 

beta dose. The other GSD values are based on the assumption that for indoor exposure the 

95th-percentile value is 10 times the median.

4.2 TONAWANDA LABORATORY BETA AND GAMMA EXPOSURE

For external dose reconstruction, workers at Tonawanda Laboratory are classified into two categories 

– research and office. The research category includes all personnel who performed hands-on work in 

research facilities (laboratories, fabrication facilities, or pilot plants) or who provided some type of 

support for these facilities that involved working in or very close to them. It includes scientists;

technicians; and shop, maintenance, stores, and janitorial personnel. The office category includes all 

personnel who had primarily desk jobs although they might occasionally have visited the research 

facilities. It includes secretaries, clerks, draftspersons, and high-level managers. Three periods 

(R&D, cleanup, and postcleanup) are considered for Tonawanda Laboratory external dose 

reconstruction. Data on worker doses were not available for Tonawanda Laboratory, but survey 

results indicated that radiation levels during pilot plant operations were similar to later measurements 

at the Ceramics Plant. Therefore, radiation level estimates were based to a large extent on the 

estimates for workers who did similar work at the Ceramics Plant. Table 4-21 shows the assumed 

annual radiation rates and their bases.

4.3 NEUTRON EXPOSURE

No neutron exposure measurements are available for the Linde site. Neutron production by means of 

the alpha-neutron reaction would have resulted in a relatively small neutron dose component during 

Step I conversion of African ore to purified U3O8; Step II conversion of U3O8 to UO2; and Step III 

conversion of UO2 to UF4. Neutron dose rates per gram of natural uranium in the compounds that 

were processed at Linde are presented in Table 4-22. 

The following assumptions were made about the conversion from neutron production rate from the 

alpha-neutron reaction to an annual dose at the receptor point: 

1. Point source geometry was used to produce a nominal ambient neutron dose rate.

2. Point source strength was based on the daily rate of U3O8 handling or UF4 production. Selfshielding within the source-target compound matrix was assumed to be negligible.

3. A 238U:235U composition of 99.3% to 0.711% by weight for natural uranium was assumed.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 60 of 102

Table 4-21. Tonawanda Laboratory beta and gamma radiation, 1942 to 1953.

Period

Job

Beta

rem/yr Gamma

GSD Basisa Start End

Hands &

forearms

Remainder

of body R/yrb GSD Basisa

R&D

10/01/1942 07/31/1946 Research 1.11E+02 3.70E+01 1.52 A 5.35E+00 2.61 F

Office 3.00E+00 3.00E+00 2.65 B 5.35E+00 2.61 B

Cleanup

08/01/1946 12/31/1946 Research 7.83E+00 2.61E+00 4.04 C 1.85E+00 4.04 C

Office 1.01E-01 1.01E-01 3.00 D 1.11E-01 4.04 D

Postcleanup

01/01/1947 12/31/1953 All 3.26E-01c 3.26E-01c (d) E 6.80E-02 (d) E

a. Basis:

A. One-half of value for most exposed Ceramics Plant process worker during 1943 to 1946 production (Step I process 

operator). The factor of 2 reduction allows for the lower intensity of research work in comparison with production. 

For example, the amount of material processed and the day-to-day hands on radiation work would have been 

significantly lower than for the Step I to III operators.

B. Based on Ceramics Plant office worker during 1943 to 1946 production.

C. Cleanup worker during Ceramics Plant decontamination.

D. Non-cleanup worker during Ceramics Plant decontamination.

E. Building 30 postcontamination level (Table 4-1) for 8.5 hr/wd plus outdoor level on Tonawanda site for 0.5 hr/wd. 

Outdoor level assumed equal to Building 30 level before vacuum cleaning and flushing (Table 4-1). A 6-day week 

was assumed.

F. Based on most exposed Ceramics Plant process worker during 1943 to 1946 production (Step I process operator).

b. Where the gamma exposure rate estimate was based on dosimetry data for the Ceramics Plant (available as values of 

DD and DM) plus possibly also outdoor dose OD, the gamma exposure rate for Tonawanda Laboratory was 

approximated as the sum of the parameters (e.g., DD + DM + OD). Gamma exposure rates are for the whole body.

c. The increased beta radiation level for office workers after the end of cleanup is due in part to allowing for the possibility 

that any worker could have been stationed anywhere in the facility once it was all considered clean.

d. Not estimated.

Table 4-22. Natural uranium per gram dose rates at 1 ft.

Chemical form

Dose ratea

(rem/hr-gram)

African ore (U3O8) with alpha emitting progeny in secular equilibrium through Ra-226 2.05E-11

Purified UO3 and UO2 with no alpha emitting progeny 7.91E-12

UF4/UF6 without alpha emitting progeny 6.62E-10

a. ORAUT-OTIB-0024 (ORAUT 2005a) data used to calculate the alpha-neutron dose rates is from Shleien, 

Slaback, and Birky (1998), Salmon and Hermann (1992), Reilly et al. (1991), and DOE (2000). 

4. The entire mass of uranium compounds was conservatively assumed to be attributable to the 

uranium (source) content. 

5. Average neutron energy from alpha-neutron reactions was 2 MeV. 

6. Dose equivalent rate-to-fluence rate conversion factor for 2 MeV was 1.3 x 10-4 rem/hr per 

neutron/cm2

-s. 

7. The work year consisted of 8 hr/d, 6 d/wk, and 50 wk/yr.

During the Step I and Step II conversion processes in Building 30 (Ceramics Plant) the bounding 

neutron dose rate would be attributable to operations that involved handling of African Ore with alphaemitting progeny in secular equilibrium through 226Ra. This arises from the fact that the African ore 

feed material contained alpha-emitting progeny that contributed to the alpha-neutron production rate. 

As indicated in Table 4-22, the African ore form produced neutron dose rates that are 2.6 times 

greater than the purified UO3 and UO2 with no alpha-emitting progeny. In addition, during Step I 

processing, water, a neutron moderator, was added to the digestion process after ore milling. 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 61 of 102

Similarly, during Step II processing, uranyl nitrate hexahydrate was added to ether, an organic 

hydrocarbon, and the solution was washed with water. The presence of these neutron moderators in 

the Step I and Step II processes would decrease the average neutron energy and consequently 

reduce the dose equivalent rate-to-fluence rate conversion factor. As a consequence, the dose that 

was received by loaders (or “Movemen”) would be bounding for all personnel who were engaged in 

Step I and Step II processing activities. As indicated in the Table 4-22, during the Step III conversion 

process in Building 38, the neutron dose rate from UF4 is almost 85 times greater than from the UO2

feed material. Therefore, the dose contribution from handling UO2 feed material can be neglected. 

As discussed in Section 2.3.2, the Linde Ceramics Plant (Building 30) processed approximately 

26,000 MT (26 million kilograms) of ore over a period of 37 months beginning in June 1943 and 

ending in July 1946. Assuming a 6-day workweek for 50 wk/yr, and a uniform rate of ore processing, 

28,111 kg/wd (5,622 kg of uranium) were handled at Linde. The assumption favorable to claimants

that all of the ore was African ore has been made although 30% of the ore was known to be 

preprocessed and would contain significantly less alpha-emitting progeny. The assumption favorable 

to claimants that all of the ore was 20% U3O8 by weight has also been made although the Linde site 

literature indicates that the ore contained 3% to 20% U3O8 by weight. According to the available 

Linde literature, the highest weekly production rate (Step III) of UF4 at the Linde Ceramics Plant 

(Building 38) was 41,624 pounds (Kent 1949a) or 3,147 kg/d.

The only potentially significant source of neutron exposure at Linde would have been neutrons 

produced by the alpha-neutron reaction in materials where uranium was mixed with elements of low 

atomic number such as fluorine and oxygen. Ceramics Plant (Buildings 30 and 38) personnel are 

assumed to be exposed to one-tenth of the daily production amount of U3O8 or UF4 at a distance of 

1 ft. This quantity of material is an estimate that is favorable to claimants of the time-averaged 

amount of material likely to have been close to the maximally exposed worker during a work shift. 

The factor of 10 reduction takes into account several factors: that the plant operated around the clock 

so that each shift dealt with only one-third of a day’s throughput; that many workers were involved in 

each type of operation so that each worked closely with only a portion of a shift’s throughput; that an 

individual who worked in the vicinity of a large quantity of material (e.g., barrels of finished product) 

would on the average have been much more than 1 ft away because of the large volume it would 

have occupied; and that even a worker in a job that involved being close to large quantities of material 

also had other activities at larger distances from the source term. 

Major uranium-related projects at Tonawanda Laboratory were as follows:

• R&D for Steps I, II and III

•

. This work occurred primarily from October 1942 to December 1943 

(Jenness and Ewing 1943). The largest-scale efforts were pilot plant projects. The Step I pilot 

plant produced 2 to 3 t of U3O8 per week in November 1942 (Bonsib 1942). The Step III 

equipment could convert up to 90 lb of UO2 to UF4 in a single run (LAPC 1946c, p. 84).

Conversion of UF6 to UO3

•

. This work appears to have been done after November 1943 

(because it is not mentioned in the progress report for work from October 1942 to November 

1943; Jenness and Ewing 1943). Interim reports were issued from July 1944 through July 

1945, and a final report was issued in May 1946 (LAPC 1946d, p. 111). Typical runs involved 

use of a 150-lb cylinder of UF6 and produced a few pounds of product. 

Grinding UF4. The laboratory conducted research on grinding UF4 at an unknown time 

between October 1942 and May 1946 (LAPC 1946c, p. 89). The work involved processing 

2,750 lb of UF4. The description of the program suggests that its duration was short (a few 

weeks).
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Based on the above and allowing for date uncertainties and the storage of materials before and after 

processing or research, the following model was used to estimate neutron doses at Tonawanda 

Laboratory:

• Ore containing 3 t of U3O8 was processed per week, which corresponds to a daily processing 

rate of 454 kg from October 1, 1942, through February 29, 1944.

• Exposures were to 100 lb of UF4 or UF6 per week daily from October 1, 1942, to July 31, 1946. 

This accounts for the Step III and UF6 to UO3 work.

• Exposures were to 3,000 lb of UF4 for 2 months at some point between October 1, 1942, and 

July 31, 1946. This accounts for the UF4 grinding research. The neutron dose in a year is 

equivalent to that which would be produced by continuous exposure to 500 lb of UF4 for a 

year. Because the dates of the work are uncertain, it was assumed that there was continuous 

exposure to 500 lb of UF4 from October 1, 1942, to July 31, 1946.

To reflect all types of UF4 and UF6 work, a daily source term of 600 lb (272 kg) was assumed. All 

source terms (both oxide and fluoride) were divided by 3 to account for the fact that no worker would 

be within close range of the source material at all times. This yielded a daily processing rate of 

151 kg of U3O8 and a continuous exposure to 90.7 kg of UF4 or UF6. Tonawanda Laboratory 

(Building 14) personnel were assumed to be exposed to these source amounts at a distance of 1 ft. 

Table 4-23 summarizes the model parameters and presents calculated dose rates. 

Table 4-23. Neutron dose rates and annual doses from the alpha-neutron sources.

Location/time period Process

Alpha 

source

Target 

atom

Daily 

source 

term (g)

Dose 

rate at 1 

foot 

(rem/hr)

Dose 

rate 

(rem/yr)

Building 30

06/01/1943–07/31/1946

Steps I and II African ore 

20% U by 

weight

Oxygen 5.62E+6 1.15E-5 2.77E-2

Building 38

07/25/1943–08/31/1946a

09/15/1947–09/30/1949a

Step III UF4 Fluorine 2.76E+6 2.08E-4 5.0E-1

Tonawanda Laboratory

(Building 14)

10/01/1942–02/29/1944

Steps I and II U3O8 Oxygen 1.51E+5 3.10E-6 7.44E-3

Tonawanda Laboratory

(Building 14)

10/01/1942–07/31/1946

Step III;

UF4 grinding;

UF6 process gas to 

UO3

UF4 or UF6 Fluorine 9.07E+4 6.01E-5 1.44E-1

a. The period of exposure was extended beyond the end of the production period to account for inventory left on hand. 

UF4 that was produced at Linde was shipped to Electromet. The dates when shipping of all Linde inventory was 

completed were estimated from data in a Linde progress report (Kent 1949b) and data on production at Electromet 

(DOE undated, pp. 23-24).

4.4 EXTERNAL DOSE RECONSTRUCTION SUMMARY, OCTOBER 1, 1942, TO 

JULY 7, 1954

This section summarizes guidelines and parameters for reconstruction of doses due to external 

radiation during October 1, 1942, to July 7, 1954 except for doses due to occupational medical 

exposure, which are addressed in Section 5.0. The parameters that are provided in this section are 

for use when individual worker data are unavailable or inadequate. Because of the SEC 

determination (HHS 2005) that it is infeasible to adequately reconstruct internal dose during the period 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 63 of 102

October 1, 1942 through October 31, 1947, dose estimates for this period are considered partial dose 

estimates.

For dose reconstruction, exposures for different jobs have been grouped into categories. Table 4-24 

specifies the categories and the annual radiation doses or exposures assigned to each. For the 

production periods at the Ceramics Plant, categories were designated high, medium, and low. In 

other cases, designations more descriptive of the work activity are used.

Table 4-24. Summary of annual external exposure from AWE operations, 1942 to 1953.

Yeara Work category

Beta (rem)a

Gamma (R)a Neutron (rem)a,b Hands & forearms Rest of body

Ceramics Plant (Buildings 30, 31, 37, and 38)

1942c All workers 2.55E–02 2.55E–02 4.97E–03 (d)

1943e High 1.51E+02f 5.05E+01 3.65E+00 3.41E–01g

5.05E+01h

Medium 1.20E+01 3.97E+00

Low 2.08E+00 2.08E+00

1944

1945

High 2.21E+02f 7.40E+01 5.35E+00 5.00E–01g

7.40E+01h

Medium 1.76E+01 5.85E+00

Low 3.00E+00 3.00E+00

1946i High 1.28E+02f 4.32E+01 3.11E+00 3.33E–01g

4.32E+01h

Medium 1.04E+01 3.59E+00

Low 1.93E+00 1.93E+00

1947i Medium 2.04E+00 8.91E–01 5.37E–01 1.48E–01g

Low 6.10E–01 6.10E–01 2.03E–01

1948 Medium 5.85E+00 1.95E+00 1.61E+00 5.00E–01g

Low 1.00E+00 1.00E+00 4.80E–01

1949 Medium/lowk 6.85E+00 2.28E+00 1.73E+00 3.74E–01g

Cleanup (d) Non-cleanupl 6.60E–01 6.60E–01 2.94E–01

1950

1951

1952

1953

1954m

Cleanupn 7.83E+00 2.61E+00 1.85E+00

(d) Non-cleanupo 3.26E–01 3.26E–01 1.11E–01

Tonawanda Laboratory (Building 14)

1942c Research 2.80E+01 9.33E+00 1.35E+00 3.63E–02

Office 7.56E–01 7.56E–01

1943 Research 1.11E+02 3.70E+01 5.35E+00 1.44E–01

Office 3.00E+00 3.00E+00

1944

1945

Research 1.11E+02 3.70E+01 5.35E+00 1.44E–01

Office 3.00E+00 3.00E+00

1946 Researchp 6.78E+01 2.26E+01 3.88E+00 8.36E–02q

Officer 1.78E+00 1.78E+00 3.15E+00

1947

1948

1949

1950

1951

1952

1953

All workers 3.26E–01 3.26E–01 6.80E–02 (d)

a. Total annual exposure (dose) for the designated year. Prorated based on calendar year and applicable notations below.

b. Because of the possible difficulty in determining whether a worker was working with oxide or fluoride materials, each 

worker was assigned the larger neutron dose due to fluorides.
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c. Exposure for the period from October 1 through December 31, 1942 only.

d. Neutron dose rate was negligible.

e. Values prorated: For January 1 through April 26, 1943 (preproduction period), applicable values from Table 4-1 applied; 

for April 27 to December 31, 1943, applicable 1944 to 1945 values applied. 

(Example calculation: 1943 high-ball mill operator = 0.315 × 1.01E-01 + 0.685 × 2.21E+02 = 1.51E+02).

f. Based on 221 rem/yr for ball mill operator, Step I and Step II process operators, and weighmaster.

g. The Building 38 neutron dose rate for Step III processing was assumed to apply from April 27, 1943, to August 31, 1946,

and from September 15, 1946, to September 30, 1949. The neutron dose rate was negligible from September 1, 1946,

to September 14, 1949 (standby), and after September 30, 1949 (cleanup and postcleanup). The period of neutron 

exposure extended beyond the end of production in 1946 and 1949 due to remaining inventory of UF4.

h. Based on 74 rem/yr for loader per Section 4.1.2.1.3.

i. Values prorated: For January 1 to July 31, 1946, applicable 1944 to 1945 values applied; for August 1 to December 31, 

1946 (standby period), applicable values from Table 4-4 (guard) applied.

j. Values prorated: For January 1 to September 14, 1947 (standby period), applicable values from Table 4-4 (guard) 

applied; for September 15 to December 31, 1947, applicable 1948 values applied. 

k. Values prorated: For January 1 to June 30, 1949 (Step III production), applicable 1948 medium values applied; for 

July 1 to December 31, 1949, 1950 to 1953 cleanup values applied.

l. Values prorated: For January 1 to June 30, 1949 (Step III production), applicable 1948 low values applied; for July 1 to 

December 31, 1949, 1950 to 1953 non-cleanup values applied.

m. Exposure for the period of January 1 to July 7, 1954 only.

n. All cleanup workers and cleanup support workers as defined in Section 4.1.5 are assigned to the cleanup exposure 

category. Parameters are those of the cleanup worker for a 6-day week in Table 4-20.

o. All non-cleanup workers as defined in Section 4.1.5 are assigned to the non-cleanup exposure category. Parameters 

are those of the non-cleanup worker for a 6-day week in Table 4-20.

p. Values prorated: For January 1 to July 31, 1946, applicable 1944 to 1945 values applied; for August 1 to December 31,

1946, applicable values from Table 4-21 (cleanup-R&D scenario) applied.

q. Includes neutron exposures through July 31, 1946.

r. Values prorated: For January 1 to July 31, 1946, applicable 1944 to 1945 values applied; for August 1 to December 31, 

1946, applicable values from Table 4-21 (cleanup-office scenario) applied.

Where the dose or exposure estimates were based on dosimetry parameters (dosimeter dose DD and 

missed dose DM) plus possible outdoor dose OD, the dose was estimated as the sum of the 

parameters (DD +DM +OD). Each estimate is considered to be the median of a lognormal distribution. 

A GSD of 3 was assigned to all beta and gamma dose estimates based on the typical GSD levels 

estimated for the underlying data; that is, beta and gamma doses were assigned as lognormal 

distributions with GSDs of 3 and an acute exposure rate. Neutron doses were assigned as a constant 

distribution and a chronic exposure rate.

Prorating of exposures over less than a year is not straightforward for all periods. If doses are top be 

prorated, the footnotes to Table 4-24 should be checked to determine how the dose should be 

distributed throughout the year. Beta doses were assigned as electrons with energy >15 keV. 

Gamma doses were assigned as photons with energy from 30 to 250 keV, which is favorable to 

claimants. Neutrons are assigned as neutrons of 0.1- to 2-MeV energy.

Table 4-25 shows how Ceramics Plant production jobs were assigned to the various categories. For 

1943 to 1946, the assignments were based on Table 4-25 and the discussion in Section 4.1.2.1.3. 

For 1947 to 1949, the assignments were based on Tables 4-16 and 4-18 and the discussion in 

Section 4.1.4.2. In each case, some job titles that reflect subcategories of titles in the earlier tables 

were added.

For the Ceramics Plant cleanup period, the cleanup category includes both cleanup workers and 

cleanup support workers as defined in Section 4.1.5. For Tonawanda Laboratory, the research and 

office categories are defined in Section 4.2.

If the exact job of a worker is not listed, dose reconstruction should be based on the most similar job.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 65 of 102

Table 4-25. Ceramics Plant worker beta and gamma external exposure categories for production 

periods

1943–1946 production 1947–1949 production

Job

Beta

category Job

Beta and gamma

categorya

Ball mill operator High A operator Medium

Chemist Medium B operator Medium

Draftsman Low C operator Medium

Engineer Medium Chemist Medium

Fire inspector Low Fire marshal Low

Guard Low First aid nurse Low

Janitor Medium Foreman, labor Medium

Lab technician Medium Foreman, loader Medium

Laundry worker Medium Foreman, maintenanceb Medium

Loader/moveman High Foreman, step III Medium

Maintenance workerb Medium Group leader, maintenanceb Medium

Nickel operator Low Group leader, step III Medium

Nurse Low Guard Low

Office workerc Low Janitor step III Medium

Ore sampler Medium Janitor, outside step III Low

Plant superintendent, asst supt Low Lab technician Medium

Seamster, seamstress Medium Laundry man Medium

Shipping & receiving clerk Low Loader/moveman Medium

Step I process operator High Maintenanceb Medium

Step II process operator High Office employee, AEC Low

Step III process operator Medium Office employee, Lindec Low

Storekeeper/stock clerk Low Shipping & receiving clerk Low

Tank farm operator Low Storekeeper/stock clerk Low

Tool crib attendant Medium Timekeeper Low

Truck driver/operator High Tool crib attendant Medium

Weighmaster High Truck driver/operator Medium

Weighmaster Medium

a. The category assignments are the same for beta and gamma radiation.

b. Maintenance personnel include all craftspersons who constructed, fabricated, repaired, or refurbished (e.g., carpenter, 

electrician, instrument repairman, lathe operator, mason, millwright, oiler, painter, pipe fitter, sheet metal worker, trades

helper, and welder).

c. Office workers include accountant, bookkeeper, clerk, comptometer operator, draftsman, industrial relations 

representative, mail person, plant superintendent and assistant superintendent, product accountant, property 

department worker, secretary, security agent, stenographer, telephone operator, and typist.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 66 of 102

5.0

Because of the SEC determination (HHS 2005) that it is infeasible to adequately reconstruct internal 

dose during the period October 1, 1942 through October 31, 1947, dose estimates for this period are 

considered partial dose estimates.

OCCUPATIONAL MEDICAL EXPOSURE 

5.1 CERAMICS PLANT EMPLOYEES

Preemployment, periodic, and termination chest X-rays were required for at least some of the workers 

at Ceramics Plant. On certain occasions, pelvis X-rays were also required.

5.1.1

The earliest record that was found of a requirement for chest X-rays at Linde is in a report of a 

conference on worker safety at Linde held March 31, 1943 (Brimm and Neumann 1943). The 

conference report called for a preemployment medical examination that included fluoroscopic chest 

examination. A September 4, 1943, letter from the Medical Section of the MED recommended initial 

and periodic examinations that included chest X-rays (Van Horn 1943a). Available medical records 

indicate that chest X-rays were regularly administered to workers in 1943 and later years. Based on 

the correspondence and the indications of plant practice found in the medical records, the 

assumptions below are provided for use in reconstruction of doses due to chest X-rays.

Chest X-rays

Ceramics Plant X-ray Assumptions

1. Applicability: These parameters apply to all

2.

job categories at the Ceramics Plant.

X-Ray Category and Frequency

a.

: 

Preemployment

b.

: Taken before starting employment at Linde Ceramics. These X-rays 

were also required for Linde employees who transferred to the Ceramics Plant from other 

parts of the company.

(1) Before August 13, 1943, taken at 3-month intervals.

Periodic

(2) From August 13, 1943, through December 31, 1944, taken at 6-month intervals.

(3) From January 1, 1945, on, taken at 12-month intervals.

c. Termination

3.

: Taken at termination of employee participation in MED/AEC work at Linde 

(either due to employee leaving or the work ending).

Methodology

a. Before July 16, 1943, by fluoroscopy of the lungs.

:

b. On or after July 16, 1943, by taking a 14- by 17-in. radiographic film.

4. Period

Bases of Assumptions

: The chest X-ray program was assumed to have started on November 16, 1942, and 

ended on December 31, 1953.

1. Applicability: A MED letter (Van Horn 1943a) recommended chest X-rays for “those persons 

working in direct contact with special materials" (Van Horn 1943a), where the term "special 

materials" appears to have denoted materials that contained uranium or uranium and radium 

(Ferry 1944b). When the medical examination program at the Ceramics Plant was 
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established, it was decided that all employees should receive the same physical examinations 

and rechecks (Cranch 1944b). By January 1944, after reports from the early examinations did 

not indicate any noticeable medical hazard, discussions were under way on reducing the 

frequency of examinations for those clerical workers who had no occasion to go into operating 

areas (Cranch 1944b). The suggestion was made that the examination frequency be reduced 

for office personnel working in the office annex building but not for certain clerical help 

(described as "some stenographers and clerks, stores clerks, etc."), who worked in the main 

building and were "from time to time exposed to the same hazards as operating personnel" 

(Holmes 1944a). A reduced frequency for guards was also suggested. By November of 1944, 

all employees were to have a preemployment examination, which included a chest X-ray. 

Employees "working in the plant or who have occasion to come in contact with any portion of 

the process or the materials" were given periodic examinations and a termination examination. 

However, employees "not having occasion to contact the process in any way" were exempt 

from the periodic and termination examinations (Holmes 1944b). From the foregoing, it is 

clear that some employees were not subject to periodic and termination X-rays. However, it is 

not clear how those employees can be identified. Therefore, the general assumption for dose 

reconstruction is that all employees were subject to the same chest X-ray requirements.

2. X-Ray Category and Frequency

3.

: Preemployment, periodic, and termination X-rays are 

discussed in various 1943 correspondence. A 3-month periodic interval was stipulated in the 

report of March 31, 1943, conference on Linde Step II safety (Brimm and Neuman 1943). A 6-

month interval was specified in a letter that was distributed August 13, 1943 (Cranch 1943). 

No medical records that exemplified a 3-month interval were found, but the available medical 

records for 1943 appear to be incomplete. Available worker files in which the first indicated 

chest X-ray date was July 1943 or later are consistent with a periodic interval of approximately 

6 months for X-rays in 1944 and a periodic interval of 12 months after 1944. 

X-Ray Type

(a) Some of the claimant files contain data sheets from the Black Rock Clinic, where Linde 

X-rays were taken, that state the size to be 14- by 17-in. (e.g., Claim File number redacted, 

DOE Response number redacted, p. 61).

: As noted, the March 31, 1943, safety conference stipulated that the 

preemployment exam should include fluoroscopic chest examination. A medical record 

indicates chest examination by "fluoroscope, lungs" in a preemployment examination on 

March 31, 1943 (Claim File number redacted, DOE Response number redacted, p. 23). A 

comment in the notes of a safety conference on May 7, 1943, indicates that the fluoroscopic 

examinations did not involve the use of film. The conference participants recommended the 

use of X-ray film to have a permanent record of the chest condition (Brimm 1943b). The 

earliest chest X-ray found in available worker files was a radiographic film from July 16, 1943 

(Claim File number redacted, DOE Response number redacted, p. 28). The type of film Xrays that were taken at Linde (radiographic rather than photofluorographic) and their size (14-

by 17-in.) was determined from several pieces of evidence:

(b) Approximately 60 arbitrarily selected Linde Air Products X-ray films in the DOE Oak Ridge 

Office vaults were pulled and viewed. All were 14- by 17-in. (Beck 2004).

(c) In 1949, Linde requested proposals for microfilming various records on cards and X-ray

films. A response to this request indicated that 3,000 14- by 17-in. X-ray films were 

involved and mentions no other size (Medhurst 1949).

One piece of evidence that suggests the use of photofluorographic X-ray techniques was 

found, a purchase requisition from November 4, 1944, for 3 gross (432) of 4.5- by 10-in. 

Type F X-ray film (LAPC 1944b). The size of this film was appropriate for photofluorographic 
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stereoscopic chest X-rays (Cardarelli et al. 2002, p. 492). However, the following factors 

suggest it was not used for such X-rays:

(1) No photofluorographic film images were found in the Linde claim files.

(2) The purchase requisition was issued by Linde and called for delivery of the film to the 

"Laboratory." If the film had been for chest X-rays, it probably would have been issued by 

the Black Rock Clinic – the only location for taking X-rays that was indicated in the Linde 

medical files that were examined – and delivered there.

(3) The purchase requisition stated the film was for "research on army contract #W-7405 Eng. 

26." This contract supported research at the Tonawanda Laboratory (Brimm and Schubert 

1945b), which suggests that the film was for use in such research. It is unlikely that the 

laboratory had its own worker X-ray program because other correspondence indicates a 

common program for the Ceramics Plant and the Proving Laboratory, the portion of 

Tonawanda Laboratory that was involved in developing uranium processing techniques 

(Cranch 1943).

4. Period

5.1.2

: November 16, 1942, is the date on which Linde entered into contract W-7401 eng-14 

for its work for the MED (Section 2.1.2). It is likely that the medical examination requirements 

were not implemented until some time after this date, but exactly how long after is not known. 

A letter to the Linde Safety Steering Committee on August 13, 1943, indicated that 

preemployment examinations were already a regular practice (Cranch 1943). The earliest 

document found that referred to the X-ray requirements was the report of the March 31, 1943,

safety conference, but the program might have been established before that date. Production 

work at the Ceramics Plant is assumed to have ended on June 30, 1949; cleanup work is 

assumed to have ended on December 31, 1953. It is likely that employees who participated in 

the cleanup work were required to have chest X-rays. In the absence of specific information, it 

is assumed that other employees were also required to have chest X-rays.

A letter from the MED Medical Section to Linde on September 4, 1943, recommended initial and 

annual pelvis X-rays for workers "who will be exposed to fluorine or fluorine compounds" (Van Horn 

1943a). However, on January 11, 1944, the MED recommended that pelvis X-rays be omitted from 

both preemployment and followup examinations (Warren 1944). To assess company practice, Linde 

medical records were checked. The review was limited in its scope because only a small portion of 

the records for the period were available – records of EEOICPA claimants and records in general 

Linde files – because of other compensation claims that had been filed against the company. Eight 

records were found that showed X-rays during the period from September 4, 1943, January 11, 1944. 

The records indicated that all of the subjects had chest X-rays, but only one also had a pelvis X-ray

(Various authors 1950, p. 67). The name of the worker for whom the pelvis X-ray was indicated was 

found in a Ceramics Plant employee list from April 1944 (Neuman 1944). The list indicated that the 

employee worked in Step II, an activity that did not involve handling of fluorine or fluorine compounds.

Pelvis/Lumbar Spine X-rays

Additional information was found in Linde correspondence files. When the recommendation to end 

pelvis X-rays was relayed to the Ceramics Plant, the plant administration advised the corporate office 

“that we have never made any pelvic X-rays of any of the people working in the Ceramics Plant” 

(Holmes 1944c). The corporate office checked with the medical doctor in its Industrial Toxicology 

Department who had oversight over medical procedures at the Ceramics Plant. The doctor wrote:

It is my understanding that no arrangement had as yet been made for the taking of 

pelvic x-rays of employees at the Ceramics plant. When this was discussed some time 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 69 of 102

ago it was decided that in view of the possible fluorine exposure it might be advisable 

later, perhaps at the WCX plant, to include pelvic x-rays when our facilities would make 

this more feasible. For this reason, a space was provided in the form for “laboratory 

examinations.” ... The opinion at present is that there would be no need at the 

Ceramics Plant to include pelvic x-rays in the physical examinations. (Cranch 1944c)

These records indicate that Linde did not institute routine pelvis X-ray examination of its employees in 

response to the September 4, 1943 recommendation from the MED. The indication of a single pelvis 

X-ray in the medical records might have related to a special situation for that employee. While dose 

reconstruction should include the X-ray examinations performed for screening in the particular 

employees medical record, this particular record is not interpreted as evidence that Linde instituted a 

general policy of taking pelvis X-rays of a particular category of employees in response to the MED 

letter (Van Horn 1943a).

In the review of employee medical records, it was noticed that the record of a female employee had 

the notations “LMP 2/10/44” and “LMP 3/4/44” in the row on the report form for recording pelvis X-ray

results. Other pages of the employee’s record had similar monthly entries, but they were in a section 

that began in the succeeding row of the report that was designated for recording results of “Other 

Tests.” This was interpreted as indicating that the two entries on the pelvis X-ray row had been 

entered on that row in error. The entries were interpreted as recording the date of the last monthly 

period in compliance with suggestions made by the MED (Van Horn 1943b). 

Pelvis X-rays of Linde employees were taken at a later time. The Ceramics Plant progress report for 

the week ending December 21, 1947 stated, "Complying with instructions received from the health 

authorities of the AEC, pelvis X-rays are being taken of all Ceramics Plant employees. It is expected 

that this program will be completed before the end of the year" (LAPC 1948). Employee medical 

records confirm that X-rays of the pelvis were taken. Images of X-ray films with the records indicate 

that 14- by 17-in. films were taken and that the long direction of the film was parallel to the length of 

the body, an orientation like that for standard lumbar spine X-rays and different from the 17- by 14-in.

orientation (the 14-in. dimension parallel to the length of the body) for standard pelvis X-rays. In each 

of the available files that had copies of pelvis X-ray films, there were two images of the pelvis. It was 

difficult to determine whether these were images of two portions of one X-ray exposure or whether 

two X-ray exposures had been taken at one session. To avoid underestimation of worker doses, for 

dose reconstruction it was assumed that two exposures were taken. 

In June 1949 as the plant was nearing shutdown, it wrote the AEC to ask whether a pelvis X-ray

should be included in the termination medical examination of all workers whether or not they had 

access to contaminated areas (Kent 1949c). A statement of the decision in this matter was not found. 

Employee medical records indicated that X-rays of the pelvis were taken at termination (around the 

end of June 1949), but the number of records was not sufficient to indicate whether they were taken of 

all employees. The X-ray film images available had characteristics similar to those from the end of 

1947.

Based on the foregoing, the assumptions below are provided for use in reconstruction of doses due to 

X-rays of the pelvis.

Assumptions

1. Applicability: These parameters apply to all

2.

job categories at the Ceramics Plant.

Occasions

a.

: Pelvis X-rays were taken at the following times:

Near the end of 1947: A pelvis X-ray was taken of every Ceramics Plant employee in 

November or December 1947.
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b. At employee termination:

3.

 A pelvis X-ray was taken of every Ceramics Plant employee who 

terminated after November or December 1947.

Number

4.

: Two exposures were taken at each occasion.

Methodology

5.1.3

: A 14- by 17-in. radiographic film was taken (17 in. parallel to the length of the 

body), a format similar to that for lumbar spine X-rays.

Dose reconstruction should consider information specific to the subject to the extent that it is 

available, adequate, and is representative of X-ray screening examinations covered under the 

EEOICPA (i.e., dose from X-ray examinations conducted as a result of occupational injuries are not to 

be included in dose reconstructions). The guidelines in this section are for use when the records for 

an individual worker are not available or are incomplete. The guidelines are for use only to the extent 

that they are not inconsistent with the worker's records. For example, if the medical records are 

complete and indicate a lower or higher examination frequency than stated in the assumptions 

provided above, the data in the medical records should be used.

X-Ray Dose Reconstruction Guidelines

X-ray doses shall be determined in accordance with the latest revision of the project technical 

information bulletin Dose Reconstruction from Occupationally Related Diagnostic X-Ray Procedures

(current version is ORAUT 2005b) when applicable. Before using the guidelines therein, dose 

reconstructors shall confirm that the assumptions on which they are based are applicable.

In relation to the above, attention is called to the following:

• For many employees, medical records are available that show the dates and types of the Xrays taken.

• Often, images of the actual X-rays are also available in claimant files.

• Some chest X-rays in Linde claimant files show portions of the chin. This indicates that 

positioning, and perhaps also collimation, was not as good as in current practice.

• X-rays of the pelvis thus far examined in Linde claimant files had the format of lumbar spine Xrays. The lumbar spine X-ray doses in the X-ray technical information bulletin might be more 

suitable for these cases than the pelvis X-ray doses.

5.2 TONAWANDA LABORATORY EMPLOYEES

The medical X-ray requirements for Ceramics Plant employees appear to have also applied to 

Tonawanda Laboratory employees who worked in pilot plant projects (Cranch 1943; Brimm 1943c). 

The extent to which the requirements were extended to other Tonawanda Laboratory employees is 

not known. In the absence of this information and in view of the likely difficulty of establishing whether 

or not a particular worker was involved in handling radioactive materials, it is assumed for dose 

reconstruction that chest X-rays were required of every Tonawanda Laboratory employee in the 

period of MED work unless it can be shown that the employee was not involved in MED projects.

The following assumptions are made about the frequency of X-rays of Tonawanda Laboratory 

employees:

1. Applicability: These parameters apply to all job categories at Tonawanda Laboratory.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 71 of 102

2. X-ray Category and Frequency

a.

: 

Initial X-ray

b.

when the worker started working on an MED contract.

Periodic X-rays

c.

at the same frequency as specified for Linde Ceramics.

Termination X-ray

3.

at termination of employee participation in MED/AEC work at Linde.

Methodology:

4.

 Same as for the Ceramics Plant.

Period:

Dose reconstruction guidelines are the same as for Ceramics Plant employees. 

 MED-related work at Tonawanda Laboratory was assumed to have started on 

October 1, 1942, and ended on December 31, 1946.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 72 of 102

6.0

This section develops parameters for reconstruction of doses due to internal and external exposures 

at the Ceramics Plant starting July 8, 1954 and Tonawanda Laboratory starting January 1, 1954.

ESTIMATION OF EXPOSURES FROM RESIDUAL CONTAMINATION AFTER 1953

Initial cleanup of Tonawanda Laboratory was assumed to be complete on December 31, 1946. 

Tonawanda Laboratory worker radiation exposures from January 1, 1947, to December 31, 1953, are 

discussed in Sections 3.0 and 4.0. 

It was assumed that beginning on January 1, 1954, Tonawanda Laboratory employees could have 

been exposed to residual contamination for 2,000 hr/yr. 

6.1 INTERNAL EXPOSURE

Airborne Radioactivity

In recent years, outdoor air radioactivity has been measured in conjunction with the site soil 

remediation program (USACE 2004). Table 6-1 displays the highest monthly average air 

concentrations of 226Ra, 230Th, and 238U for July 2000 through June 2004. Assuming that the 

measured 238U is part of a natural uranium source term, the total uranium concentration would be the 

238U concentration divided by the fraction of activity due to 238U (0.4886), which is 4.3 × 10-4 pCi/m3

. 

The activity ratios of 230Th and 226Ra to uranium are 0.84 and 1.7, respectively. 

Table 6-1. Highest monthly outdoor airborne radionuclide 

concentrations at Linde Tonawanda site, 2000 to 2004.

Nuclide Highest monthly concentration (pCi/m3)

a

Ra-226 7.5E-04

Th-230 3.6E-04

U-238 2.1E-04

a. Per analysis of data from USACE (2004).

Building 30 was found in 1976 to be the most contaminated building on the site (ORNL 1978). The 

indoor airborne uranium concentration in 1976 was measured as 1.9 × 10-2 pCi U/m3 (ORNL 1978, 

Table 4-2). This larger uranium air concentration was used to estimate a chronic intake. To maintain 

the consistent intake units and estimate the annual intake, the uranium air concentration of 

1.9 × 10-

2 pCi/m3 was multiplied by 2.22 dpm/pCi, a breathing rate of 1.2 m3

/hr, and 2,000 hr/yr. The 

annual intake rate was estimated at 100 dpm/yr for all Tonawanda employees beginning in 1955. 

Dividing by 365 d/yr gives a daily uranium inhalation intake rate of 0.277 dpm/d. The 230Th and 226Ra 

daily inhalation intake rates would be 0.233 and 0.471 dpm/d, respectively. Ingestion intake rates 

were estimated using the steps in Section 3.7. The summarized intake results are shown in 

Table 6-2. It is unlikely that uranium would be in a chemical form consistent with absorption type F 

during the residual contamination period, so it was assumed that only types M and S would be inhaled 

during this period. Thorium-230 could be type M or S, and 226

Ra is type M. 

Workers noted that Building 30 renovations occurred in the 1960s that could have influenced air 

concentrations. Specific details of the renovation, including the actual period of renovation, dust 

control measures, location of work and occupancy of areas, are not available. It is reasonable to 

assume that renovations could have resulted in elevated airborne radioactivity. Because a maximum 

value of measured air concentration was used to estimate the intakes during the residual exposure 

period, a factor of 14 was assumed to be reasonable to describe the uncertainty at the 95th percentile 

that is associated with the possibility of elevated intakes during Building 30 renovations. So for organ 

doses, the distribution was assumed to be lognormal with a GSD of 5.

Other progeny are partially accounted for by the assigned radon exposure. 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 73 of 102

Radon Progeny

Fifty-five measurements of radon progeny concentration were made in 1976 and 1981 in the 

Tonawanda site buildings that were used in MED/AEC work (ORNL 1978, pp. 17 and 84; BNI 1982, 

p. B-24). The 1981 survey was more comprehensive and yielded significantly higher concentrations, 

so the 1976 results were ignored. Building 31 had the highest radon progeny concentration. The GM

of the PAEC was 1.68 × 10-2 WL with a GSD of 1.89 (based on analysis of data in BNI 1982, 

Table B-3). No correction was made for natural background radon, because its value was unknown. 

Based on 12 work months per year, the radon exposure was estimated as 2.01 × 10-1 WLM/yr, and is 

listed in summary Table 6-2 in the next section.

6.2 EXTERNAL BETA AND GAMMA EXPOSURE

As noted, Building 30 was found in 1976 to be the most contaminated building on the site (ORNL 

1978). Measured floor and wall radiation levels in Building 30 in 1976 were compared with similar 

measurements from 1950 immediately after its decontamination (see Section 4.1.1 and Heatherton 

1950). The results were similar, but the 1950 values were slightly higher. 

Outdoor gamma radiation levels at 1 m above the surface on and near the Tonawanda site were 

measured in 1976 and 1981. Natural background gamma radiation levels in the Tonawanda area 

were said to be 8 to 15 μR/hr (ORNL 1978, p. 15; BNI 1982, p. B-12). ORNL 1978 (p. 15) reported 

only isolated areas with levels significantly higher than background; readings in these areas were 

described as very nonuniform and varying from background up to 250 μR/hr. Regions with readings 

above about 20 μR/hr were indicated in a figure in the report (Fig. 26). Figure B-5 of BNI (1982)

summarized the combined results of the two studies. Locations and values were shown for readings 

≥25 μR/hr. BNI (1982, p. B-12) stated that the points of maximum radiation that were determined in 

the ground surveys were slightly displaced from but in general agreement with those that were 

determined in a 1979 aerial radiological survey. The total number of reported readings ≥25 μR/hr was 

16. The net readings (after subtraction of 8 μR/hr to correct for background) had a GM of 94 μR/hr

and a GSD of 3.95. This was taken as an estimate of worker exposure rate when outdoors. This 

estimate was assumed to apply starting January 1, 1954, at the Tonawanda Laboratory and July 8, 

1954, at the Ceramics Plant. No credit was taken for the soil remediation at the site that began in 

2000 because the remediation has not been completed (Pilon 2004). The estimate is probably an 

overestimate for the following reasons: measurements were reported only for areas that were 

identified as having above-background radiation levels, these areas occupied much less than half of 

the open area of the site (based on Fig. B-5 of BNI 1982), and the reported measurements were not 

averages over these areas but only readings at a few hot spots. 

The beta dose rate at 3 ft above the ground that corresponded to the adopted value of gamma 

exposure rate was estimated as 4.38 × 10-1 mrem/hr, which is 4.66 times the gamma rate where 4.66 

is the ratio of beta mrem/hr at 3 ft to gamma mR/hr at 3 ft for the floors and walls in Building 30 after 

decontamination (Table 4-1). It is assumed that the GSD for the beta dose rate would be the same as 

for the photon dose rate.

Because the radiation levels seemed to remain fairly constant, and because the levels were fairly low, 

the Tonawanda Laboratory postcleanup exposures, based on Building 30 contamination levels, were 

used to estimate the external exposure rate. No adjustments for changes in work hours were made. 

Table 6-2 summarizes the results. The radiation energy distributions were assumed to be same as 

those during the operational period.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 74 of 102

Table 6-2. Annual internal and external exposure to residual radioactivity.

Internal

Start/End 

Datea Exposure Absorption type Intake (dpm/d) IREP distribution

U-234 (b) Inhalation M, S 2.77E-01 Lognormal GSD 5

(b) Ingestion (c) 5.78E-03 Lognormal GSD 5

Th-230 (b) Inhalation M, S 2.32E-01 Lognormal GSD 5

(b) Ingestion (c) 4.84E-03 Lognormal GSD 5

Ra-226 (b) Inhalation M 4.71E-01 Lognormal GSD 5

(b) Ingestion (c) 1.01E-02 Lognormal GSD 5

WLM/yr

Rn-222 (b) Inhalation - 2.01E-01 Lognormal GSD 1.89

External Start Exposure Basis

(b) Penetrating Survey instrument 0.068 R/yr Lognormal GSD 3

(b) Non-penetrating Survey instrument 0.326 rem/yr Lognormal GSD 3

a. January 1, 1954, start date applies to Tonawanda Laboratory; for the Ceramics Plant, the higher dose values that are 

associated with D&D operations (Sections 3.0 and 4.0) should be applied until July 7, 1954, and the values cited in this 

table apply starting on July 8, 1954, and should be prorated accordingly.

b. The exposure end data has been designated as July 2006 (Turcic 2007). 

c. Choose same f1-value as for inhalation per NIOSH (2004).

6.3 NEUTRON EXPOSURE

Because only small quantities of radioactive material were on the site in the postcleanup period, there 

were no significant neutron exposures.

7.0

All information requiring identification was addressed via references integrated into the reference 

section of this document.

ATTRIBUTIONS AND ANNOTATIONS
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 75 of 102

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Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 86 of 102

GLOSSARY

carnotite 

Mineral of the form K2(UO2)2(VO2)2·

3

(H2O) found in Colorado and Utah and mined for its 

uranium and vanadium content. Much of the domestic ore processed at Linde Ceramics Plant 

derived from tailings from carnotite ore. Preprocessing of this material before shipment is 

believed to have removed much of its radium content (Aerospace Corporation 1981).

day (d)

As used in this Linde document, a day refers to a calendar day unless otherwise described. A 

calendar day is any 24-hour day in a year. There are at least 365 calendar days in a year. A 

workday (wd) refers to a day that is assigned for work. The length of a workday depends on 

the amount of time spent on the job. Default assumptions are that there are 8 hours in a 

workday and 250 workdays in a year. At Linde, the number of hours per workday and the 

number of workdays per year for some employees were higher than the defaults for the early 

years.

Linde database 

Set of records of Linde Ceramics Plant operations available to the National Institute for 

Occupational Safety and Health dose reconstruction project.

N.G. or NG Cake 

Filter cake produced in Step II processing at Linde. There were two types of cakes produced 

called OK and N.G.

pitchblende 

Mineral containing uranium oxide (UOX) of variable composition ranging between UO2 and 

U3O8 as well as radium. Raw ore from Africa was processed for its uranium content at Linde 

Ceramics Plant the raw ore was not prerefined and so contained all of the natural uranium 

radioactive progeny including radium.

Step I, Step II, and Step III 

Sequential uranium production processes at Linde: 

• Step I, conversion of ore to U3O8 (black oxide)

• Step II, conversion of U3O8 to UO3 (orange oxide) to UO2 (brown oxide)

• Step III, conversion of UO2 to UF4 (green salts)

torbernite 

Mineral in the form of Cu(UO2)2(PO4)2·8-12(H2O) processed for its uranium content. Like the 

African pitchblende processed at Linde, torbernite was not prerefined before its arrival at Linde 

and so contained all of the natural uranium radioactive progeny including radium.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 87 of 102

ATTACHMENT A

BETA RADIATION

Page 1 of 4

LIST OF TABLES

TABLE TITLE

A-1 Maximum beta surface dose rates from various uranium-containing materials.......................

PAGE

87

A-2 Electron energy released by refined uranium ......................................................................... 87

A-3 Electron energy released by unrefined natural uranium ore ................................................... 88

A-4 Estimated attenuation of beta radiation with distance............................................................. 89

A-5 Uranium beta dose reduction factors for apparel.................................................................... 89

A-6 Variation of dose rate with body location for Step III process operations................................ 90

This attachment contains data that was used in analyzing exposures of workers to beta radiation.

Table A-1. Maximum beta surface dose 

rates from various uranium-containing 

materials.

Source

Beta surface dose rate

mrad/hr

a,b rad/8 hr

Slab of U metal 233 1.86

UO2 207 1.66

UO3 204 1.63

U3O8 203 1.62

UF4 179 1.43

a. From Table 2-7 on p. 2-18 of DOE (2000).

b. Beta surface dose rate in air through a 

polystyrene filter 7 mg/cm2 thick.

Table A-2. Electron energy released by refined uranium.

Principal 

nuclides

Electron energy

per decay of nuclide

(MeV/nt)a

Electron energy

per decay of parent

(MeV/nt)a % of total

U-238 series U-238 0.010 0.005 1.1

Th-234 0.060 0.029 6.6

Pa-234m 0.822 0.402 89.8

U-234 0.013 0.006 1.4

U-235 series U-235 0.049 0.001 0.2

Th-231 0.165 0.004 0.8

Total: 0.447

a. nt = nuclear transformation.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 88 of 102

ATTACHMENT A

BETA RADIATION

Page 2 of 4

Table A-3. Electron energy released by unrefined natural uranium ore.

Principal 

nuclides

Electron energy

per decay of nuclide

(MeV/nt)a

Electron energy

per decay of parent

(MeV/nt)a % of total

U-238 series U-238 0.010 0.005 0.4

Th-234 0.060 0.029 2.5

Pa-234m 0.822 0.402 34.7

U-234 0.013 0.006 0.5

Th-230 0.015 0.007 0.6

Ra-226 0.004 0.002 0.2

Rn-222 0.000 0.000 0.0

Po-218 0.000 0.000 0.0

Pb-214 0.293 0.143 12.4

Bi-214 0.659 0.322 27.8

Po-214 0.000 0.000 0.0

Pb-210 0.038 0.019 1.6

Bi-210 0.389 0.190 16.4

Po-210 0.000 0.000 0.0

U-235 series U-235 0.049 0.001 0.1

Th-231 0.165 0.004 0.3

Pa-231 0.065 0.001 0.1

Ac-227 0.016 0.000 0.0

Th-227 0.053 0.001 0.1

Ra-223 0.076 0.002 0.1

Rn-219 0.006 0.000 0.0

Po-215 0.000 0.000 0.0

Pb-211 0.456 0.010 0.9

Bi-211 0.010 0.000 0.0

Tl-207 0.493 0.011 1.0

Total: 1.157

a. nt = nuclear transformation.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 89 of 102

ATTACHMENT A

BETA RADIATION

Page 3 of 4

Table A-4. Estimated attenuation of beta radiation with distance.

Distance from

surface (m)

Relative dose ratea

Refined natural

uraniumb

Unrefined natural uranium 

ore or ore byproductsc

0.0 1.00 1.00

0.3 0.80 0.88

0.94 0.50 0.66

1.0 0.48 0.64

1.6 0.31 0.50

2.0 0.23 0.42

3.0 0.11 0.27

5.0 0.025 0.11

8.5d 0.000 0.024

10.0 (e) 0.012

12.6f (e) 0.000

a. Only exponential attenuation due to air is considered. Additional reduction 

would come from geometric dispersion, which is ignored here.

b. Dose rate assumed to vary with distance as exp(-0.0074x), where x is 

distance from the surface in centimeters. This was derived from a fit to data 

in Coleman, Hudson, and Plato (1983).

c. Calculated falloff rate based on the assumption that all of the beta radiation is 

due to the 3.27-MeV beta ray from Bi-214. The dose rate is assumed to vary 

with distance as exp(-0.0044x), where x is distance from the surface in cm. 

The attenuation factor 0.0044/cm was estimated from the rule of thumb that 

the half thickness of a beta absorber is one-eighth of the range of the beta 

rays (Cember 1983, p. 97). For natural uranium ore and byproducts with beta 

emitters other than Bi-214, the actual falloff with distance would be greater 

because all other significant beta emissions have lower energies than 

3.28 MeV.

d. Range of 2.28-MeV beta radiation per Cember (1983, p. 99).

e. No radiation at this distance.

f. Range of 3.27-MeV beta radiation per Cember (1983, p. 100).

Table A-5. Uranium beta dose reduction factors 

for apparel.

Itema Fraction of betadose remaining

Leather, medium weight 0.62

White cotton gloves 0.89

"Tyvek" coveralls 0.98

65% Dacron/35% cotton lab coat 0.91

a. Selected from Table 6-11 on p. 6-23 of DOE (2000).
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 90 of 102

ATTACHMENT A

BETA RADIATION

Page 4 of 4

Table A-6. Variation of dose rate with body location for Step III process operations.

Operation

(mR/hr)a

Hands, bare Hands, gloved Face Chest Legs Feet

Handling empty trays 4 3 1.5 4 (b) (b)

Loading trays (brown) 25 13 2 4.5 4.5 2

Handling loaded trays(brown) 25 13 5 16 (b) (b)

Handling loaded trays (green) 25 16 7.5 25 (b) (b)

Unloading operation (green) 25 (b) 7.5 7.5 4 2

Blending 15 10 7 8.5 12 (b)

Radiation level relative to chestc

Radiation level 

relative to hands, 

barec

Hands, bare Face Chest Legs Feet Hands, gloved

Handling empty trays 1.00 0.38 1.00 (b) (b) 0.75

Loading trays (brown) 5.56 0.44 1.00 1.00 0.44 0.52

Handling loaded trays(brown) 1.56 0.31 1.00 (b) (b) 0.52

Handling loaded trays (green) 1.00 0.30 1.00 (b) (b) 0.64

Unloading operation (green) 3.33 1.00 1.00 0.53 0.27 (b)

Blending 1.76 0.82 1.00 1.41 (b) 0.67

Maximum 5.56 1.00 1.00 1.41 0.44

a. Data from Heatherton (1948h).

b. No data.

c. Results in lower half of the table are based on the data in the upper half.
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 91 of 102

ATTACHMENT B

CODES AND SPECIAL TERMINOLOGY

Special codes and terminology were used in correspondence and documents that related to Linde's 

work for the MED and AEC in place of common names or to specify special forms of materials with 

particular characteristics. Over the course of time, some terms might have had multiple meanings. 

Table B-1 lists apparent definitions deduced during preparation of this document. 

Table B-1. Codes and special terminology.

Code or term Apparent meaning

C-103 UO2 from DuPont; a feed material used for Linde Step III

C-306 UO2 from Mallinckrodt; a feed material used for Linde Step III

C-316 UO2 from Mallinckrodt; a feed material used for Linde Step III

F-29 U3O8 made at Linde

K-25 UF6 gas containing various concentrations of U-235

K-65 Residues containing Ra and Pb that were byproducts of the ore to UO3 conversion 

process at Mallinckrodt Chemical Works

L-19 A domestic ore processed at Linde; estimated to contain mostly 10% to 16.5% 

U3O8 (Aerospace 1981, Table B-1)

L-30 African pitchblende ore; estimated to contain 8% to 12% U3O8 (Aerospace 1981, 

Table B-1)

L-50 African pitchblende ore; estimated to contain 6.7% U3O8 (Aerospace 1981, 

Table B-1)

Mx, MX Uranium or U3O8

MX-308 U3O8 (in ore)

My Radium

Mz, MZ Radon

O-71 UF4

P-65 UO2 produced by Linde Step II; a feed material used for Linde Step III

Product 65 P-65

Q-20 Torbernite ore processed at Linde; estimated to contain 17.7% U3O8 (Aerospace 

1981, Table B-1)

R-10 African pitchblende ore processed at Linde; estimated to contain 3.5% U3O8

(Aerospace 1981, Table B-1)

T Uranium

Tubealloy dioxide UO2

Tubealloy tetrafluoride UF4

X Uranium or uranium ore 
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 92 of 102

ATTACHMENT C

DATA SOURCES ON URANIUM PROGENY CONCENTRATIONS IN LINDE MATERIALS

calculated value

<MDA, mda shown

SRDB Ref ID Date Location Description U238 U234 U235 totalU Th230 Ra226 Th232 Th228 Ac227 Pa231 units Th230/U Ra226/U Th232/U Ac/U Pa/U

16294 1995 bldg 14 dry valve pit (dust) 1038 1068 53.8 2159.8 354.7 15.3 2.3 1.7 pCi/g 0.16 0.01 0.00

sump (dust) 1.4 1.7 0.07 3.17 0.72 0.52 0.2 0.22 pCi/g 0.23 0.16 0.06

corridor overhead (dust) 369 378.8 11.6 759.4 60.2 0.14 5 0.67 pCi/g 0.08 0.00 0.01

corridor wall (terra cotta block) 3.2 3.2 0.14 6.54 1.7 1.4 1.4 1.4 pCi/g 0.26 0.21 0.21

14620 1978 bldg 30 air samples during D&D 1.90E-08 2.10E-09 1.10E-09 pCi/ml 0.11 0.06 0.00

9009 1981 sediment, onsite and offsite Ellicott Creek 0.82 0.05 1.69 0.6 0.55 0.7 pCi/g 0.36 0.33 0.41

Creek1 0.95 0.05 1.95 0.7 0.7 0.8 pCi/g 0.36 0.36 0.41

Twomile Creek - upstream 4.3 0.1 8.7 0.92 0.69 0.01 pCi/g 0.11 0.08 0.00

Twomile Creek - linde discharge 0.71 0.06 1.48 0.02 0.52 0.02 pCi/g 0.01 0.35 0.01

Twomile Creek - downstream 1.5 0.05 3.05 0.96 0.59 0.48 pCi/g 0.31 0.19 0.16

Storm Sewer 6.47 0.19 13.13 1.4 1.35 0.62 pCi/g 0.11 0.10 0.05

Storm Sewer 99 4.57 202.57 18 6.93 0.51 pCi/g 0.09 0.03 0.00

Storm Sewer 13 0.52 26.52 2 1.59 0.65 pCi/g 0.08 0.06 0.02

Storm Sewer 116 4.1 236.1 9.9 0.89 0.34 pCi/g 0.04 0.00 0.00

Storm Sewer 4.5 0.17 9.17 0.2 0.64 0.39 pCi/g 0.02 0.07 0.04

Sanitary Sewer 362 13 737 1.33 1.94 0.11 pCi/g 0.00 0.00 0.00

Sanitary Sewer 0.51 0.05 1.07 0.34 0.38 0.21 pCi/g 0.32 0.36 0.20

9026 1990 Linde Soils area 1 - mean 11.2 22.4 7.8 4.3 1.6 pCi/g 0.35 0.19 0.07

area 2 - mean 12.7 25.4 5.7 3.4 1.4 pCi/g 0.22 0.13 0.06

area 3 - mean 17.1 34.2 24.4 9.4 1.4 pCi/g 0.71 0.27 0.04

area 4 - mean 46.8 93.6 30.7 9.8 1.4 pCi/g 0.33 0.10 0.01

8828 1981 Linde Soil and Sediment near disposal well - subsurface (loc 11) 24.05 0.84 48.94 5.9 5.53 0.92 14.25 0.73 pCi/g 0.12 0.11 0.02 0.29 0.01

test well debris (loc 13) 26.4 1.09 53.89 3.53 0.82 0.51 2.1 0.29 pCi/g 0.07 0.02 0.01 0.04 0.01

Sanitary Sewer (loc 15) 362 12.93 736.93 1.33 1.14 0.11 5.54 0.95 pCi/g 0.00 0.00 0.00 0.01 0.00

Storm Sewer (loc 19) 99.2 4.57 202.97 17.7 6.93 0.51 14.29 1.14 pCi/g 0.09 0.03 0.00 0.07 0.01

Storm Sewer (loc 21) 116 4.1 236.1 3.89 0.89 0.34 3.07 0.39 pCi/g 0.02 0.00 0.00 0.01 0.00
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 93 of 102

LIST OF TABLES

TABLE TITLE

D-1 Linde data sources in the Site Research Database................................................................

PAGE

94

D-2 Uranium urinary excretion data .............................................................................................. 95

D-3 Chronic intake rates for Type M 234U ...................................................................................... 96

D-4 Chronic intake rates for Type S 234U....................................................................................... 96

LIST OF FIGURES

FIGURE TITLE

D-1 Type M 50th-percentile value for November 1, 1947, to January 27, 1950.............................

PAGE

96

D-2 Type S 50th-percentile value for November 1, 1947, to January 27, 1950 ............................. 96

D-3 Type M 84th-percentile value for November 1, 1947, to January 27, 1950............................. 97

D-4 Type S 84th-percentile value for November 1, 1947, to January 27, 1950 ............................. 97

Due to the limited availability of bioassay data from the Linde site, it was necessary to conduct a 

coworker study of all the bioassay data for use to determine intake estimates. The data that were 

used in this study were transcribed directly from hardcopy into worksheets. The data in the 

worksheets were verified (as indicated below), a statistical analysis was conducted and verified, and 

an intake assessment was conducted and verified. Each of these processes is further described 

below. The resultant intake tables are provided in the intake assessment section. 

Data Verification

The Linde bioassay data were verified as follows:

1. Data were transcribed directly from source documents to spreadsheets by data entry 

personnel. A review of the datasets was conducted to determine the datasets in each source 

document (see Table D-1). The data were present in sampling lots (e.g., a group of samples 

related by sample date).

2. A single, master data spreadsheet was created by merging spreadsheets from individual 

source documents.

3. A completeness and accuracy review of 100% of the data was conducted by an individual 

independent of spreadsheet creation. Any discrepancies in data entry were identified and 

corrected. Two samples in the December 1950 dataset were excluded from further analysis 

because they were identified as belonging to employees at the electrometallurgical plant (i.e., 

not Ceramics Plant employees). One sample in the November 1948 dataset was excluded 

because it was a duplicate (same name and result).

ATTACHMENT D

LINDE URANIUM COWORKER ASSESSMENT FOR NOVEMBER 1947 TO JANUARY 1950

Page 1 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 94 of 102

Table D-1. Linde data sources in the Site Research Database (SRDB).

Period

Samples

(count)

SRDB Ref ID

8955 11148 16979

Nov-1947 (11/11–11/12) 73 X

Jan-1948 (01/04–01/06) 88 X X Xa

Feb-1948 (02/20–02/23) 84 X X

June-1948b (06/17–06/19) 69 X X

Aug-1948 (08/18) 67 X X

Nov-1948 (11/12–11/17) 94 X X

Feb-1949 (02/07–02/10) 91 X Xc

Apr-1949 (04/11) 4 X

Jul-1949 (07/01–07/14) 88 Xd

Sep-1949 (09/11–09/12) 2 X

Oct-1949 (10/20) 1 X

Dec-1949 (12/30) 10 X

Jan-1950 (01/23–01/27) 9 X

680

a. Pages 97–98 and 696–697 only.

b. Datasheet in SRDB Ref ID 11148 version has handwritten correction of sampling month 

(from June to May).

c. Dataset is incomplete.

d. Samples are annotated as “termination”.

4. Dataset completeness was evaluated through a review of available data on the Linde 

urinalysis program.

a. The urine data are from transmittals from the laboratory to Linde. Based on a review of 

each individual dataset, and after referencing was conducted between datasets (using the 

sample identifiers that did not appear to vary from dataset to dataset), it appeared that 

each individual dataset was complete.

b. A handwritten listing of Linde urine datasets (ca. 1980) was found in the SRDB (Author 

unknown undated). Based on this listing, one dataset (October 1, 1948) is currently not 

available.

c. The transmittal letter for the February 1948 urine samples (March 4, 1948), clarifies that 

Linde was to submit quarterly urine samples (instead of monthly, as was previously 

understood). The frequency of the available Linde urine data are consistent with this 

pattern (i.e., monthly sampling up to February 1948 and quarterly thereafter).

With the exception of the October 1948 dataset, all known urine data have been included in this study.

The verified data were analyzed in accordance with the requirements in ORAUT-PROC-0095, 

Generating Summary Statistics for Coworker Bioassay Data (ORAUT 2006). 

Statistical Analysis

1. The 73 samples collected in November 1947 that were annotated as preemployment were not 

used in the coworker analysis because preemployment samples would not relate to exposures 

on the job.

ATTACHMENT D

LINDE URANIUM COWORKER ASSESSMENT FOR NOVEMBER 1947 TO JANUARY 1950

Page 2 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 95 of 102

2. Seven followup samples (four in April 1949, two in September 1949, and one in October 1949) 

were excluded from statistical analysis. These followup samples indicated a reduction in 

urinary excretion in every case. Exclusion of these samples is favorable to claimants.

3. Two samples from December 1949 that were labeled ”Area Plant - Electrometallurgical” were 

excluded from analysis because they were not Ceramics Plant employees.

4. One sample from November 1948 was a duplicate sample entry (typographical error) and 

therefore was excluded from further data analysis.

5. To perform statistical analysis, samples were grouped based on sampling date. Because

most of the data were clustered around sampling campaigns, the midpoint of the sampling 

interval for each sampling event was generally assigned as the effective bioassay date. The 

effective bioassay dates are used in the Integrated Modules for Bioassay Analysis (IMBA) 

software to calculate the intake rates. 

6. The uranium mass excretion rates were converted to the uranium activity excretion rates by 

applying the specific activity of natural uranium (683 pCi/mg) and urination rate of 1.4 L/d. 

7. The analytical limit of detection was assumed to be 0.01 mg/L. This value is based on 

annotations on the urine data transmittal memoranda that accompanied most datasets. 

Because none of the data were explicitly labeled as <LOD, all nonzero data were included in 

the data ranking and the curve fitting.

8. A lognormal distribution was assumed for the urinary excretion data, and the 50th- and 84thpercentile uranium excretion rates were calculated using the method in ORAUT-PROC-0095

(ORAUT 2006). These excretion rates are listed in Table D-2. 

Table D-2. Uranium urinary excretion data (pCi/d).

Period

Number of 

samples

Effective

bioassay date

50th percentile

(pCi/d)

84th percentile

(pCi/d)

01/04/48–01/06/48 88 01/05/48 20.4 62.0

02/20/48–02/23/48 84 02/21/48 15.8 47.2

06/17/48–06/19/48 69 06/18/48 9.5 15.9

08/18/48 67 8/18/48 4.9 14.1

11/12/48–11/17/48 93 11/14/48 6.5 26.1

02/07/49–02/10/49 91 02/08/49 2.9 14.9

07/01/49–07/14/49 88 07/07/49 10.9 46.4

12/30/49–01/27/50 17 01/13/50 4.9 15.3

597

Intake Assessment

Urine results in Table D-2 were used in the intake assessment. Samples from between December 

1949 and January 1950 were not used because the July 1949 dataset was labeled as termination 

samples for the bulk of the Linde employees. All urinary excretion rates were modeled as normally

distributed 24-hour urine samples with a uniform relative error of 30%. The excretion data were 

modeled with IMBA for multiple chronic intakes of Type M or Type S uranium. 

ATTACHMENT D

LINDE URANIUM COWORKER ASSESSMENT FOR NOVEMBER 1947 TO JANUARY 1950

Page 3 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 96 of 102

The intake rates, GSDs, and periods to which they apply are listed in Table D-3 for Type M uranium, 

and in Table D-4 for Type S uranium. 

Table D-3. Chronic intake rates for Type M 234U.

Start date End date

50th-percentile 

value (pCi/d)

84th-percentile 

value (pCi/d) GSD

11/01/1947 01/27/1950 74 297 4.0

Table D-4. Chronic intake rates for Type S 234U.

Start date End date

50th percentile 

(pCi/d)

84th percentile 

(pCi/d) GSD

11/01/1947 01/27/1950 1,884 8,085 4.3

Plots of expected and observed urinary excretion from these fits are shown in Figures D-1 through 

D-4. In each plot, the solid line represents a fit of the coworker bioassay data. 

Figure D-1. Type M 50th-percentile value for November 1, 1947, to January 

27, 1950.

Figure D-2. Type S 50th-percentile value for November 1, 1947, to January 

27, 1950.

ATTACHMENT D

LINDE URANIUM COWORKER ASSESSMENT FOR NOVEMBER 1947 TO JANUARY 1950

Page 4 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 97 of 102

Figure D-3. Type M 84th-percentile value for November 1, 1947, to January 

27, 1950.

Figure D-4. Type S 84th-percentile value for November 1, 1947, to January 

27, 1950.

ATTACHMENT D

LINDE URANIUM COWORKER ASSESSMENT FOR NOVEMBER 1947 TO JANUARY 1950

Page 5 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 98 of 102

Original Issue

During the SC&A-conducted interview in Buffalo, Linde site experts and past workers 

indicated that there were many thousands of used burlap bags stacked up in the open 

bay area behind Building 30 (see Attachment 3 of this review report). These bags 

were used for transporting uranium ore to Linde for processing. After the end of the 

operation period, the contaminated burlap bags were stored behind Building 30 

awaiting disposal. Many Linde workers, operation staff, and administrative personnel 

sat on these contaminated bags during breaks and lunch periods. This practice 

continued for many years, exposing many people at close distances to beta and 

gamma radiation sources left over in the uranium-contaminated burlap bags. The site 

profile does not estimate the missed beta and gamma doses to workers resulting from 

sitting on or standing next to those burlap bags. (SC&A 2006)

Subsequent Analysis

During the March 26, 2007, Linde Working Group meeting the burlap bag issue was discussed by 

members of the working group, NIOSH, and S. Cohen & Associates (SC&A). The following task was 

defined by the working group (based on SC&A 2006):

NIOSH to investigate details of used burlap bags. Which bags (formerly containing 

African or domestic ore) were stored at which locations and during which periods of 

time. This may affect both internal and external exposures. Even though African ores 

were processed only during the SEC [Special Exposure Cohort] period (pre 

10/31/1947), empty bags, that had contained African ore, may have been around 

longer (i.e., after 1950).

NIOSH to determine whether there was an on-site incinerator to burn used burlap bags 

and, if so, the possible effects on internal and external exposures.

In November 2007, an analysis of the burlap bag issue was conducted by NIOSH and submitted to 

the working group along with responses to the other issues that had been identified in the site profile 

review. Table 5-1 of the response document summarizes the issue and is reproduced below (NIOSH 

2007c).

Date Data Source Data

Undated Medical Section, Manhattan 

District, Data Sheet for Industrial 

Hazard Rating

(MED undated)

• Step I material (ore) arrives in burlap bags, covered 

in paper sacks 

• Step III material (UO2) arrives in 75 lb drums

1943–1946 Linde Construction, Operations 

Report

(LAPC 1946c)

• Empty bags and drums are held until they can be 

burned or scrapped

• Bags from dry ore are shaken in the bag shaker ... 

the bags are next taken to the laundry room for 

washing. All bags are then burned in the incinerator. 

Ashes from the incinerator are returned to the 

grinding circuit for processing.

ATTACHMENT E

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD

Page 1 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 99 of 102

Date Data Source Data

08-25-1943 “Safety Precautions, Ceramics 

Plant”

(Murphy 1945)

• Bags in which low-grade ore is received are to be 

sent back to the original supplier

• Bags in which African ore is delivered are to be 

burned to recover the elements

07-04-1944 “Interview with [Name redacted; 

Privacy Act Information], a Former 

Linde Employee,” April 10, 1981

(Dupree 1981)

• Washing of burlap bags in which the ore was 

received began

07-15-1944 Step I - L-30 Ore

(Olevitch 1944)

• Empty ore bags contain about 1⁄2 lb of ore

• Washing burlap bags recovers 70% of material 

• After washing, the bags are removed, centrifuged, 

and stored out in the backyard of the plant. To date 

about 19,000 bags have been washed and stored.

10-10-1944 “Interview with [Name redacted; 

Privacy Act Information], a Former 

Linde Employee,” April 10, 1981

(Dupree 1981)

• Burning of burlap bags in which the ore was received 

began

10-20-1944 “Visit to the Tonawanda Area, 18 

and 19 October 1944”

(Tybout 1944a)

• The contractor has decided not to use the bag 

washer for the bags from L-50 (the low grade ore)

• The incinerator which will be used for burning the 

bags was inspected. At the time, it had been in 

operation only one or two days.

10-??-1944 Interview with [Name redacted; 

Privacy Act Information], a former 

Linde employee, April 10, 1981

(Dupree 1981)

• The burlap bags had to be disposed of. To control 

the ore dust that remained in the bags after they were 

dumped and prevent dust from escaping during their 

disposal, a series of mechanical devices were 

installed. First a bag shaker was added to the 

disposal process to shake out ore dust left behind 

after dumping. Than a bag washer was installed to 

wash any material remaining in the bag after it had 

been shaken. Finally, a bag incinerator was installed 

to burn the bags after they had been washed. This 

entire process was completely installed by October, 

1944.

02-22-1945 E. L. Van Horn (USACE) to A. R. 

Holmes (LAPC)

(Van Horn 1945)

• ... It is desired, however, to continue film monitoring 

on employees engaged in burning L-30 ore bags ... 

06-20-1946 “Interview with [Name redacted; 

Privacy Act Information], a Former 

Linde Employee,” April 10, 1981

(Dupree 1981)

• Burning of burlap bags in which the ore was received 

ends

Just before the January 8 working group meeting, a review of NIOSH (2007c) from SC&A was 

received (SC&A 2008):

Section 5.0 of the NIOSH response discusses the “burlap bag issue.” NIOSH 

concludes that “Based on the reviewed historical records, and considering the fact that 

the period during which the burlap bags were staged and burned is within the current 

SEC period, a revision to the current dose reconstruction methodology is not 

warranted” (NIOSH 2007[c], Section 5.0). This, however, does not adequately respond 

ATTACHMENT E

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD

Page 2 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 100 of 102

to the site expert interview assertion that thousands of burlap bags were still stacked 

behind Building 30 after 1950; as stated in the SC&A site profile review: “During the 

MED period, they stacked all the contaminated burlap bags in storage area of 

Building 30. These contaminated burlap bags were kept in there until they were 

removed to be burned in the incinerator in the late 1950s. Many of the people working 

in Building 30, including operation personnel, secretaries, and maintenance workers, 

would sit on those bags resting or eating their lunch. This went on for many years.” 

(SC&A 2006, Attach. 3, p. 112). 

During the January 8 working group meeting an attempt was made to resolve the burlap bag issue, 

considering both NIOSH (2007c) and SC&A (2006). A consensus on the subject could not be 

reached. The conclusion of the working group was to have a technical conference that focused on 

the burlap bag issue. SC&A agreed to review the interview notes that were associated with the issue. 

As a result, a conference call was held on February 13, 2008. The former employee that reported the 

burlap bag issue was able to participate on this call.

Based on clarification of the burlap bag issue from the former employee who made the original 

assertion, it was determined that the burlap bags in question were inside Building 30:

... there were two pallets of bags containing some material in Building 30 in the time 

period around August 1951. He and his friends would get drinks, stand by the pile of 

bags, and place their cups on the bags. He was told by one co-worker that the bags 

were left over from the Manhattan project. He was asked about the appearance of the 

bags and he replied that they looked like large canvas sandbags.

An affidavit that describes the burlap bags was submitted by this former employee along with the 

Linde SEC Petition Documents that were submitted on March 19, 2008.

Based on this newly clarified description of the burlap bags, NIOSH was tasked with an assessment 

to compare calculated doses from an assumed pile of bags (two pallets of bags 2 to 3 ft high, for both 

ore and concentrate) to currently assigned doses. 

Comparison of currently assigned dose to potential exposure from ore bags

1.85 R/yr (GM) with GSD of 4.04 from ORAUT-TKBS-0025 (ORAUT 2005c), Table 18.

Currently assigned dose (1950 to 1954)

(note: 95th-percentile value would be 18.5 R/yr).

ATTACHMENT E

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD

Page 3 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 101 of 102

Measurements take in 1944 on African ore (Skinner 1944)

Dose assessment for the maximizing case:

Location Gamma dose rate

Ore sampling room

15 ore bags in center on ore 0.019 R/hr

1 ft from ore bags 0.0062 R/hr

Ore box car

Center of car on ore 0.017 R/hr

Ore storage area (pile 4 by 5 by 20 ft)

Center of 4- by 20-ft face on ore 0.040 R/hr

Center of 4- by 20-ft face on ore 0.026 R/hr

1 ft from 4- by 20-ft face 0.013 R/hr

2 ft from 4- by 20-ft face 0.013 R/hr

3 ft from 4- by 20-ft face 0.010 R/hr

5 ft from 4- by 20-ft face 0.0061 R/hr

10ft from 4- by 20-ft face 0.0028 R/hr

Based on the description of the exposure conditions from the former employee, the following 

exposure conditions and assumptions can be inferred:

• A dose rate of 0.0062 R/hr based on 1 ft distance from 15 ore bags, and 

• An exposure duration of 1 hr/wd, 250 wd/yr.

These result in a total annual exposure of 1.5 R/yr

Summary

Based on the weight of the available evidence (tabulated below), it is unlikely that two pallets of 

uranium ore (which was last processed at Linde in 1946) would have been in Building 30 in 1951 

(5 years after the cessation of processing of uranium ore). The current external exposure model for 

the period in question incorporates uncertainty in the external dose assignment by application of a 

lognormal distribution with a GM of 1.85 and a GSD of 4.04. This assumed distribution (with a 

95th-percentile value of 18.5 R/yr) accounts for possible deviation of the actual worker exposure of 

the magnitude that would result from the assumption that two pallets of uranium ore were in 

Building 30 in 1951.

ATTACHMENT E

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD

Page 4 of 5
Document No. ORAUT-TKBS-0025 Revision No. 01 Effective Date: 11/04/2008 Page 102 of 102

Information that contradicts the presence of 

2 pallets of uranium ore in Building 30 in 1951

Information that supports the presence of 

2 pallets of uranium ore in Building 30 in 1951

1. LAPC documentation outlines the processing of 

uranium ore including the receipt of uranium ore 

in bags, handling of the ore bags, and methods 

that were used to recover material held up in 

bags after they were dumped (shaking, washing, 

incineration, and recycling of incinerator ash).

2. LAPC documentation establishes that uranium 

ore was received in bags and that ore 

concentrates were received in drums.

3. LAPC documentation establishes end of uranium 

ore processing as 1946.

4. Affidavit (1981) from a former employee 

substantiates the description of the handling of 

the uranium ore bags in the LAPC 

documentation.

5. Surveys of Building 30 in 1950 and the balance 

of the site in 1952 did not identify elevated 

radiation levels consistent with two pallets of 

uranium ore.

1. Affidavit (2008) from a former Linde employee 

identifies the presence of 2 pallets of bagged 

material in Building 30 in 1951.

2. Affidavit (2008) from a former Linde employee 

identifies the material as uranium ore based on 

statement from a coworker.

It should be noted that the uranium intake value in this revision of the site profile (15,200 dpm/d) 

would bound any internal exposure scenario from two pallets of uranium ore bags in the stated 

configuration (bagged with plastic lining) and conditions (storage).

ATTACHMENT E

FOCUSED ASSESSMENT OF DOSE CONSEQUENCES FROM URANIUM ORE BAGS ON THE 

SITE DURING THE POSTOPERATIONS PERIOD

Page 5 of 5